US2926067A - Concentration of pu using an iodate precipitate - Google Patents

Description

l'Jnited rates Patent @fifice 2,926,067 Patented Feb. 23, 1960 CONCENTRATION OF Pu USING AN IODATE PRECIPITATE Bernard A. Fries, Richland, Wash, assignor to the United States of America as represented by the United States Atomic Energy Commission No Drawing. Application July 17, 1946 Serial No. 684,247

5 Claims. (Cl. 2314.5)

This invention relates to methods for processing materials containing the element of atomionumber 94 known as plutonium, and for separating and concentrating this element. More particularly, this invention concerns the treatment of plutonium in materials initially derived from irradiated uranium to obtain plutonium in a concentrated state and of a high quality free from certain contaminants.

As described herein, the isotope of element 94 having a mass of 239 is referred to as 94 and is also called plutonium, symbol Pu. in addition, the isotope of element having a mass of 239 is referred to as 93 Ref erence herein to any of the elements is to be understood as denoting the element generically, whether in its free state or in the form of a compound, unless indicated otherwise by the context.

Elements 93 and 94 may be obtained from uranium by various processes which do not form a part of the present invention including irradiation of uranium with neutrons. Neutron irradiated uranium may be prepared by reacting uranium with neutrons from any suitable neutron source, but preferably the neutrons used are obtained from a chain reaction of neutrons with uranium.

Naturally occurring uranium contains a major portion of U a minor portion of U and small amounts of other substances such as UX and UX When a mass of such uranium is subjected to neutron irradiation, particularly with neutrons of resonance or thermal energies, U by capture of a neutron becomes U which has a half life of about 23 minutes and by beta decay becomes 93 The 93 has a half life of about 2.3 days and by beta decay becomes 94 Thus, neutron irradiated uranium contains both 93 and 94 but by storing such irradiated uranium for a suitable period of time, the 93 is converted almost entirely to 94 In addition to the above-mentioned reaction, the reaction of neutrons with fissionable nuclei such as the nucleus of U results in the production of a large number of radioactive fission products. As it is undesirable to produce a large concentration of these fission products which must,in View of their. high radioactivity,be separated from the 94 and further as the weight of radioactive fission products present in neutron irradiated uranium is proportional to the amounts of 93 and 94 formed therein, it is preferable to discontinue the irradiation of the uranium by neutrons when the combined amount of 93 9 and 94 is equal to approximately 0.02 percent by weight of the uraniumrnass. At this conentration of these substances, the concentration of fission elements which must be removed is approximately the same percentage.

As described above, there are certain extraneous materials present with 1161 11 as a result of its method of manufacture, and it is desirable to remove these materials. However, in the removal. of certain extraneous materials other contaminants may be introduced such as for example a content of iron which may be picked up from the stainless steel vessels that are used in processing the irradiated Pu containing materials. Dependent upon the type of use to which the Pu is to be employed or the derivatives to be prepared therefrom, it may be desirable to eliminate a substantial part or all of these various contaminants. In addition it is apparent that it is desirable to accomplish this production of a relatively pure Pu or derivative thereof in a minimum number of steps.

A number of processes have already been proposed for accomplishing separation and recovery of Pu. One of these processes which is in successful use is generically known as the bismuth phosphate process. Another process which is useful in treating Pu for decontamination and the recovery of Pu is known as the wet fluoride process. These processes are the invention of others and the details thereof are described in copending applications. For example, reference may be made to app. Ser. No. 519,714, now Patent No. 2,785,951, to be referred to in further detail hereinafter, which describes the aforementioned processes. Consequently, all of the details respecting such processes are not described herein.

While the aforementioned processes have been used successfully and serve to separate and concentrate Pu, it is proposed by the present invention to provide a supplemental and revised procedure which may be employed for obtaining Pu or derivatives thereof having the particular quality of being free of certain contaminants, yet employing a minimum of steps and less recycling than heretofore required. As indicated above, the plutonium-containing materials may be contaminated or become contaminated with various components such as lanthanum, iron, bismuth, and zirconium. The substances may have been initially present, or picked up from equipment during the process. For certain purposes, the presence of such contaminants are not desired. For example, the presence of iron in Pu may be disadvantageous if it is desired to prepare the peroxide derivative of Pu. The presence of iron tends to cause peroxide decomposition. Similar remarks apply to the presence of other contaminants. Consequently, the present process is particularly advantageous for use when it is desired to produce Pu, or certain derivatives thereof, free of particular contaminants and produce Pu of high purity in a simple manner With a minimum of recycling or other similar steps. 7

The meaning of the terms, bismuth phosphate process, wet fluoride process, decontamination, product or by-product precipitation and similar terms will be apparent as the description proceeds.

This invention has for an object to provide a method for the separation and recovery of plutonium.

Another object is to provide a method for the separation and recovery of plutonium wherein the plutonium or derivatives thereof may be obtained in a form relatively free of certain contaminants.

Still another object is to provide a process particularly useful in the concentration of plutonium from materials which have been at least partially decontaminated by other procedures.

Still another object is to provide a method for the separation and recovery of plutonium involving the formation of an iodate compound.

Another object is to provide a method of concentrating plutonium which lends itself to the use of and to coupling with various extraction and decontamination procedures already known or practiced.

Still another object is to provide a process for the concentration of plutonium employing less recycling and utilizing a small number of operations, comparatively, as respects the magnitude of the product yields and the quality of the Pu produced.

Still another object is to provide a process for separating and recovering relatively pure Pu or a derivative thereof from sources which contain small amounts ofcontaminants such as lanthanum and iron.

A still further object is to provide a method of treat- 3 ment which may be applied to plutonium containing materials for rendering the materials easily susceptible of subsequent treatment such as isolating the peroxide derivative thereof.

Other objects will appear hereinafter.

I have found that Pu in admixture with various extraneous components of the type referred to above may be recovered and concentrated with a high degree of success by interposing certain iodate precipitation steps in the concentration procedure. I have found that by forming an iodate derivative of the Pu that this will bring down the Pu substantially quantitatively. The lanthanum and iron contaminants will to a large extent be caused to remain in the supernatant liquid and thereby may be rendered separable from the Pu.

The general and overall operation of my process will now be considered broadly. The solution containing Pu which is to be separated and recovered may be from any usual source, such as that referred to above obtained by dissolving neutron irradiated uranium materials in a suitable solvent such as nitric acid. The materials may be subjected to one or more extractions and decontamination steps in accordance with the bismuth phosphate method. Likewise, the partially decontaminated and concentrated materials may be processed by a wet fluoride procedure such as lanthanum fluoride precipitation. The foregoing operation is in accordance with practice already known. For example, the bismuth phosphate type of extraction is described in app. Ser. No. 519,714, filed January 26, 1944, in the names of Thompson and Seaborg, now U.S. Patent 2,785,951, and reference is made to that application for full disclosure of such process, details thereof being omitted from the present disclosure except when necessary to an understanding of the present invention. As set forth in said application, it is now known that plutonium has more than one oxidation state, including a lower oxidation state or states referred to herein as Pu" in which the element is characterized by forming insoluble phosphates and fluorides and higher oxidization states referred to as Pu in which the element forms soluble phosphates and fluorides. As is further set forth in 519,714, now latent No. 2,785,951, said lower oxidation state includes Pu+ and said higher oxidation state includes plutonium in the hexavalent state. It should therefore be understood that, as used herein, Pu includes plutonium in the tetravalent state, and Pu includes plutonium in the hexavalent state.

The exact manner of carrying out the steps is not a limitation upon the present invention. It is merely preferred that the materials be processed by some method as exemplified by the bismuth phosphate and wet fluoride processes to eliminate bulky components such as uranium and to at least partially decontaminate and concentrate the Pu. That is, the Pu is isolated from a substantial part of the fission products and similar extraneous components which may be present because of the method of manufacturing the Pu by neutron irradiation as above described.

-In the present invention, the Pu containing materials such as a lanthanum fluoride product precipitate carrying Pu are placed in solution by metathesis. That is a known treatment with alkali hydroxides, or carbonates may be applied as will be described in greater detail hereinafter. Subsequently, nitric acid is added so that a lanthanum nitrate solution containing the Po is obtained. This metathesized solution is then treated with a source of iodate ions in accordance with the present invention. This causes the formation of an iodate precipitate of the Pu leaving in the supernatant liquid lanthanum and iron contaminants which it is desired to eliminate from the Pu.

The iodate precipitate is separated and redissolved. A source of hydroxyl ions is added to cause the formation of a hydroxide precipitate of the Pu. This precipitate may be readily separated and redissolved. After the elimination of contaminants such as iron and lanthanum in the preceding steps, the solution containing Pu which is obtained is susceptible of further treatments for still additional purification, if desired, and for converting the Pu to derivatives such .as the peroxide derivative. As described above, the aforementioned treatment eliminates iron' and otherwise purifies the Pu containing materials so that the formation of various derivatives may be accomplished without the possibility of such contaminants rendering the derivatives unstable and susceptible to decomposition.

Considering now the process in greater detail, an illustration of extraction and decontamination by the bismuth phosphate method is set forth below. While the uranium may be subjected to extraction and decontamination by any suitable process, a preferred process is that described in app. Ser. No. 519,714 aforementioned, now US. Patent 2,785,951, an embodiment of which is as follows. Neutron irradiated uranium is dissolved in a suitable quantity of 6070% nitric acid. This gives a solution containing Pu. The solution is treated with a reducing agent such as H 0 in excess for a period of about one hour at a temperature from 50 C. to 75 C. whereby any of the Pu which may have been oxidized to the Pu state in the solution step is reduced to the Pu" state. The concentration of the solution in the uranyl nitrate hexahydrate is adjusted to 20% and H is added to make the solution 1 N therein. To the solution is now added a source of bismuth ion to provide a concentration of bismuth ion equivalent to 10 grams of Bi+ ion in four liters of 20% uranyl nitrate hexahydrate; phosphoric acid is also added to make the solution .36 M therein, and a precipitate comprising BiPO which carries the P11" comes down and is separated from the solution by filtration or centrifugation. The BiPO precipitate carrying the Pu is dissolved in 10 N HNO The acidity of the solution is reduced to 6 N HNO by dilution and the solution made .1 M in K CI O On heating the solution at 95 C. for 2.25 hours, the plutonium is oxidized to the Pu state. The solution is then diluted to 1 N acidity by addition of Water and H PO added to provide a suitable concentration for causing the formation of a BiPO precipitate. The solution is heated toabout C. whereupon BlPOq, precipitates carrying fission products but not Pu The precipitate may be removed by filtration or centrifugation and discarded. If repetition of the cycle is contemplated for further decontamination, the Pu in the filtrate is reduced by passing in a rapid stream of S0 gas for five minutes and allowing the solution to stand for approximately one hour and the cycle is suitably repeated.

The fluoride method is now briefly described:

The plutonium-containing solution with the plutonium in the reduced state is acidified by adding nitric acid to make the solution .8 to 1.3 N therein. To this solution there is now added a source of lanthanum ion, for example, an aqueous solution of lanthanum nitrate containing about one percent lanthanum ammonium nitrate to provide a concentration of lanthanum ion equivalent to 50-250 milligrams La per liter of solution. The solution is then treated with hydrogen fluoride as 48% HF or other suitable source of fluoride ion to make the solution about .5 to 1 N therein and to cause formation of a precipitate comprising lanthanum fluoride. This lanthanum fluoride precipitate carries the plutonium in the reduced condition and may be separated by centrifugation or other procedure. In the event the fluoride treatment is conducted with Pu in an (0) state a by-product precipitate carrying contaminants occurs. The Pu remains in solution. The lanthanum fluoride treatment and its details of operation forms no part per se of the present invention as various other concentration or chemical procedures may be applied, but is merely described for illustrating certain treatments which may be used.

The metathesis referred to above is as already indicated a prior art procedure as respects this application and briefly comprises the following:

The plutonium-carrying lanthanum precipitate is treated in a suitable vessel by adding several liters of to percent KOH or NaOH solution thereto and maintaining the resulting slurry between about 50 C. and 85 C. for a periodof l to 2 hours. Exchange of hydroxyl ions for fluoride ions of the lanthanum fluoride occurs to form an insoluble lanthanum hydroxide which carries the plutonium, while the fluoride ions go into solution. The metathesized material is separated from the solution and is carefully washed with water to remove fluoride ion therefrom as completely as possible. Alternatively, the plutonium-carrying lanthanum fluoride precipitate may be metathesized with a solution of potassium carbonate. For example, 45 liters in two steps of addition of a 45 to 50 percent solution of potassium carbonate may be added to a vessel containing the lanthanum fluoride precipitate, and the resultant slurry maintained at a temperature of 90 C. to 75 C. for a period of about 1-2 hours. The lanthanum fluoride precipitate dissolves in the potassium carbonate solution. Then the resulting solution is made 15% in KOH by adding 40 percent solution of KOH or NaOI-l to form a lanthanum hydroxide precipitate which also carries the plutonium. This precipitate is separated as by centrifugation and is carefully washed to remove fluoride ion. The washed fluoride-free materials resulting from either of the foregoing metathesis procedures may be treated with dilute nitric acid or other suitable solvent or additions to yield a solution upon which my iodate precipitation may be carried out.

Referring now to the treatment of the lanthanum nitrate solution containing Pu resulting from the aforementioned steps, this solution may contain certain contaminants which it is desired to eliminate. For example, in plant operations carried out in stainless steel or other iron containing metal tanks, a content of iron may be picked up which contaminates the Pu. Or, if certain types of preliminary extraction and decontamination procedures have been employed, the Pu may contain a content of lanthanum. By adding a source of a material furnishing the iodate ion the precipitation of plutonium iodate is accomplished thereby leaving behind in the supernatant liquid at least a substantial portion of the contaminants such as lanthanum and iron.

For example, I have found that the content of iron contaminant, by means of my iodate precipitation, may be reduced by a factor of ten. That is, the amount of iron impurity which may be eliminated in the supernatant liquid is ten times that which carries along with the Pu. The iodate precipitate, so freed of contaminants, may be readily redissolved in a solvent such as nitric acid after conversion to the hydroxide, in the manner comparable to conventional practice. The resultant solution contain ing Pu is relatively free of lanthanum and iron, or, dependent upon the prior concentration of these contaminants contains a smaller amount thereof.

The HNO solution is oxidized with known oxidizing agents such as potassium dichrornate, potassium permanganate, silver nitrate and a potassum persulfate to oxidize the Pu" if it is desired to eliminate contaminants such as bismuth and zirconium. That is, substantially complete elimination of the contaminants may be accomplished by applying at this point a wet fluoride by-product precipitation, such as a lanthanum fluoride precipitation, under oxidizing conditions. This will cause the formation of a lanthanum fluoride precipitate that carries down a substantial part of the remaining bismuth and zirconium contaminants.

At this stage in the process since the volumes and contaminants are relatively minute, as compared with first stages of the process, the application of a lanthanum fluoride precipitation (or K PuF procedures) reduces the amount of La required to carry Pu. Or no La may be required and potassium plutonium fluoride may be pre- '6 cipitated. -In other words, the lanthanum additions at this stage are relatively small and are consumed in accomplishing the bismuth and zirconium removal Without the necessity of repeated precipitation or recycling for purification.

In the step of forming a K PuF precipitate this is accomplished by reducing the oxidized solution from the by-product lanthanum fluoride precipitation step, with sulfur dioxide or other reducing agent such as oxalic acid. This causes the formation of the KZPUFG precipitate which settles leaving contaminants in the supernatant. If desired, a source of potassium ions such asp'otassium nitrate may be added. However, the solutions generally contain suficie'nt residual potassium ions, from the aforementioned oxidizing agents, and residual fluoride ions so that the compound forms merely upon supplying the re ducing agent.

In some instances where it is not desired to take a LaF by-product precipitate as described, the Pu hydroxide after converting to the nitrate may be treated directly to precipitate the peroxide. The formation of the peroxide may be carried out in a conventional manner.

My process may be further illustrated by the specific examples set forth below which show the preferred embodiment of the present invention. As has been indicated, certain of the steps referred to are individually not new or are not regarded as my invention. However, the combination of steps set forth and particularly the step of forming an iodate compound of the P11 is believed to be novel.

' Example I A source of Pu which has been treated by various steps so that a lanthanum fluoride precipitate containing Pu is obtained, constiultes the starting materials, from which it is desired to obtain relatively pure Pu. In this precipitate, for example, the ratio of Pu to La may be of the order of 1 to 14.

v This lanthanum fluoride precipitate is put into solution by standard alkali hydroxide metathesis or by potassium carbonate treatment. The details of these treatments have already been described above. By the metathesis treatment together with the step of nitric acid dissolution, there is now obtained a lanthanum nitrate solution containing Pu contaminated with La and possibly some Fe.

This lanthanum solution, about 3 N in nitric acid, was made about .15 M in K10 This caused the formation of a precipitate comprising Pu(IO This plutonium iodate precipitate contained only a small amount of lanthanum. The supernatant liquid remaining contained the bulk of the lanthanum. The iodate precipitate was separated in the conventional manner by filtering.

The precipitate of plutonium iodate was redissolved in l M sulfur dioxide water.

Ammonium hydroxide was added to the solution from the preceding step to form a plutonium hydroxide (Pu(OH) precipitate. The precipitate was separated by a conventional procedure leaving the supernatant liquid containing iodide ions, resulting from the reduction of the iodate by the sulfur dioxide, and the sulfur dioxide residuum or other solvent component.

The plutonium hydroxide precipitate was then redissolved in nitric acid to give a plutonium nitrate solution. This solution may possibly contain a small amount of lanthanum. However, i211. this point in the process, the plutonium to lanthanum ratio has now been changed at least to about 14 to 1 thereby obtaining about fold concentration as respects lanthanum purification.

As discussed, the solution of plutonium nitrate may also contain a small amount of bismuth and zirconium. These various components may now be substantially completely eliminated by subjecting the plutonium nitrate to a standard lanthanum fluoride by-product precipitation under oxidizing conditions, if desired. In View of 7 the concentration factor, aforementioned, which has been obtained, this lanthanum fluoride by-product precipitation step is on a relatively small scale. The by-product precipitate carrying contaminants such as bismuth and zirconium may be separated conventionally by centrifuging or filtering and discarded.

The solution from the preceding step containing Pu is reduced with sulfur dioxide. Under reduced conditions a precipitate of K PuF is thrown down leaving any remaining contaminants in the supernatant liquor.

The potassium plutonium fluoride formed may then be separated and metathesized by a. conventional procedure with alkalis as already described and subjected to peroxide precipitation, after nitrate formation or other standard treatment, for converting the Pu to the derivatives desired for further use.

Example II In accordance with this example, the solution to be treated contained about 2 parts by weight of Pu as P-u(NO and 30 parts by weight of the contaminant La, as La(NO This solution was treated with S and NH OH to give a La(OH) and Pu(O'I-I) precipitate. A loss of only about .019% Pu in the supernatant liquid occurred.

The hydroxide precipitates, after separation by centrifugation were dissolved in HNO and the resultant solution made 3 N in HNO and 0.15 M in K10 A precipitate of Pu(IO formed. The loss in this step was about .034%.

The Pu(IO was separated and dissolved in 1 M 80;;

water. NH OH was added to the solution for precipitating Pu(OH) The loss in this step of Pu was about 064%.

The Pu(0I-I) was separated and dissolved in HNO and the solution oxidized with Ag+ and S O A content of HF Was incorporated in the oxidized solution. A LaF lay-product precipitate was formed and separated. The loss of Pu in the by-product precipitate was about 206%.

The supernatant from the preceding step containing P-u was reduced with S0 A precipitate comprising K PuF formed and was separated. The F u in this compound was relatively pure and free of contaminants such as Fe, La, Bi and Zr. The loss of Pu in this step was about .75 The overall loss in obtaining a high quality Pu material was about 1.07%.

Some of the advantages of the above procedure include: Providing an alternative method for concentrating Pu, providing a process involving less recycling because of smaller product losses and providing a process utilizing a smaller number of operations for obtaining a higher product yield.

The proportions and concentrations set forth in the above examples are illustrative and other amounts may be employed. For example, the lanthanum nitrate solution may be from 3 to N in HNO The iodate addition may be from .05 M to .15 M or higher in either K10 HlO or other similar source of iodate ion. In place of using S0 for dissolving the Pu (10 precipitate other solvents such as K C0 (30% 45% solution) may be used. When alkali carbonates are so used preferably KOH is added to precipitate the Pu(OH) It has been found that Pu(IO dissolves very readily in 30%45% K CO at room temperature. The Pu is precipitated as Pu(OH) by the addition of KOH, and IO3 is removed in the K CO -KOH supernatant with only negligible losses of Pu. In connection with the foregoing it will be noted that if S0 is used to dissolve the iodate, then Pu(OH) would be obtained; otherwise, Pu(OH) would be formed. The Pu(OH) on dissolving and heating with HNO is converted to Pu(NO It is to be understood that all matter contained in the above description and examples shall be interpreted as 8 illustrative and not limitative of the scope of this invention.

I claim:

1. In a process for the recovery of plutonium values comprising carrier precipitation of the plutonium values, in tetravalent oxidation state, on a lanthanum fluoride precipitate as the carrier, and subsequent derivation from said precipitate of an aqueous, acidic lanthanum solution containing said plutonium values in tetravalent oxidation state, by means of separating and converting said lanthanum fluoride precipitate to substantially a lanthanum hydroxide precipitate carrying said plutonium values, and thereupon separating and dissolving the resulting hydroxide precipitate in an aqueous 3 to 5 N nitric acid solution to form said lanthanum solution, the improvement step which comprises thereafter incorporating potassium iodate in said solution to make the solution 0.05 to 0.15 molar therein and thereupon precipitate said plutonium values therefrom as plutonous iodate, and separating the resulting plutonous iodate precipitate from its lanthanumcontaminated supernatant solution.

2. In a process for the recovery of plutonium values comprising carrier precipitation of the plutonium values, in tetravalent oxidation state, on a lanthanum fluoride precipitate as the carrier, and subsequent derivation from said precipitate of an aqueous, acidic lanthanum solution containing said plutonium values in tetravalent oxidation state, by means of separating and converting said lanthanum fluoride precipitate to substantially a lanthanum hydroxide precipitate carrying said plutonium values and then separating and dissolving the obtained hydroxide precipitate in an aqueous nitric acid solution to form said lanthanum solution, the improvement steps which comprise thereafter incorporating potassium iodate in said solution to precipitate said plutonium valum therefrom as plutonous iodate, separating the resulting plutonous iodate precipitate from its lanthanum-contaminated supernatant solution, deriving from said iodate precipitate an aqueous, acidic solution containing said plutonium values, and incorporating in the derived solution a source of hydroxyl ion to precipitate said plutonium values. therefrom as hydroxide.

3. In a process for the recovery of plutonium values comprising carrier precipitation of the plutonium values, in tetravalent oxidation state, upon a lanthanum fluoride precipitate as the carrier, and subsequent derivation from said precipitate of an aqueous, acidic lanthanum solu tion containing said plutonium values in tetravalent oxi-- dation state, by means of separating and converting said lanthanum fluoride precipitate to substantially a lanthanum hydroxide precipitate carrying said plutonium values and then separating and dissolving the obtained hydroxide precipitate in aqueous 3 to 5 N nitric acid to form said lanthanum solution, the improvement steps which comprise incorporating potassium iodate in said solution to render the solution 0.05 to 0.15 molar therein and thereupon precipitate said plutonium values therefrom as plutonous iodate, separating the resulting plutonous iodate precipitate from its lanthanum-contaminated supernatant solution, dissolving the separated plutonous iodate in sulfur dioxide water, and incorporating ammonium hydroxide in the resulting sulfur dioxide water solution to precipitate said plutonium values therefrom as hydroxide.

4. In a process for the recovery of plutonium values comprising carrier precipitation of the plutonium values, in tetravalent oxidation state, on a lanthanum fluoride precipitate as the carrier, and subsequent derivation from said precipitate of an aqueous, acidic lanthanum solution containing said plutonium values in tetravalent oxidation state by means of separating and converting said lanthanum fluoride precipitate to substantially a lanthanum hydroxide precipitate carrying said plutonium values and then separating and dissolving the obtained hydroxide precipitate in aqueous 3 to 5 N nitric acid to form said lanthanum solution, the improvement steps which comprise incorporating potassium iodate in said solution to render the solution 0.05 to 0.15 molar therein and thereupon precipitate said plutonium values therefrom as plutonous iodate, separating the resulting plutonous iodate precipitate from its lanthanum-contaminated supernatant solution, dissolving the separated plutonous iodate precipitate in an aqueous alkali carbonate solution, and incorporating potassium hydroxide in the resulting plutonium-containing alkali carbonate solution to precipitate said plutonium values therefrom as plutonous hydroxide.

5. In a process for the decontamination and recovery of plutonium values having bismuth and zirconium contamination associated therewith comprising carrier precipitation of the plutonium values, in tetravalent oxidation state and accompanied by said contamination, on a lanthanum fluoride precipitate as the carrier and subsequent derivation from said precipitate of an aqueous, acidic lanthanum solution containing said plutonium values in tetravalent oxidation state and contaminated with said contamination by means of separating and con verting said lanthanum fluoride precipitate to substantially a lanthanum hydroxide precipitate carrying said plutonium values and then separating and dissolving the obtained hydroxide precipitate in aqueous nitric acid to form said lanthanum solution, the improvement steps which comprise thereafter incorporating potassium iodate in said solution to precipitate plutonium values therefrom as plutonous iodate, separating the resulting plutonous iodate precipitate having associated therewith said bismuth and zirconium contamination from its lanthanumcontaminated supernatant solution, dissolving the separated plutonous iodate precipitate in substantially l M sulfur dioxide water, incorporating ammonium hydroxide in the resulting sulfur dioxide Water solution to precipitate said plutonium values therefrom as hydroxide, separating and dissolving the resulting plutonium hydroxide precipitate along with said bismuth and zirconium contamination still associated therewith in aqueous nitric acid, then while maintaining said plutonium values, so dissolved, in their hexavalent oxidation state precipitating lanthanum fluoride therein to carrier precipitate therewith said bismuth and zirconium contamination, and after removing the resulting contaminant-containing lanthanum fluoride precipitate from its plutonium-containing supernatant, reducing said plutonium values therein to their tetravalent oxidation state and precipitating them as potassium plutonous fluoride.

References Cited in thefile of this patent Friend: Textbook of Inorganic Chemistry, vol. 7, part 3, page 299 (1926); publ. by Charles Grifiin & Co., Ltd., London.

Seaborg: Chemical and Eng. News, vol. 23, No. 23, pages 2190-2l93 (December 1945).

Seaborg et al.: Journal of the American Chemical Society, vol. 70, pages 1128-4134 (1948); report submitted in 1942.

Claims (1)

1. IN A PROCESS FOR THE RECOVERY OF PLUTONIUM VALUES COMPRISING CARRIER PRECIPITATION OF THE PLUTONIUM VALUES, IN TETRAVALENT OXIDATION STATE, ON A LANTHANUM FLUORIDE PRECIPITATE AS THE CARRIER, AND SUBSEQUENT DERIVATION FROM SAID PRECIPITATE OF AN AQUEOUS, ACIDIC LANTHANUM SOLUTION CONTAINING SAID PLUTONIUM VALUES IN TETRAVALENT OXIDATION STATE, BY MEANS OF SEPARATING AND CONVERTING SAID LANTHANUM FLUORIDE PRECIPITATE TO SUBSTANTIALLY A LANTHANUM HYDROXIDE PRECIPITATE CARRYING SAID PLUTONIUM VALUES, AND HEREUPON SEPARATING AND DISSOLVING THE RESULTING HYDROXIDE PRECIPITATE IN AN AQUEOUS 3 TO 5 N NITRIC ACID SOLUTION TO FORM SAID LANTHANUM SOLUTION, THE IMPROVEMENT STEP WHICH COMPRISES THEREAFTER INCORPORATING POTASSIUM IODATE IN SAILD SOLUTION TO MAKE THE SOLUTION 0.05 TO 0.15 MOLAR THEREILN AND THEREUPON PRECIPITATE SAID PLUTONIUM VALUES THEREFROM AS PLUTONOUS IODATE, AND SEPARATING THE RESULTING PLUTONOUS IODATE PRECIPITATE FROM ITS LANTHANUMCONTAMINATED SUPERNATANT SOLUTION.