US2857241A - Process using potassium lanthanum sulfate for forming a carrier precipitate for plutonium values - Google Patents

Description

2,857,241 Patented Oct. 21, 1958 ice PROCESS USING POTASSIUM LANTHANUM' SUL- FATE FOR FORMING A CARRIER PRECIPITATE FOR PLUTONHUM VALUES No Drawing. Application October 16, 1944 7 Serial No. 558,953

Claims. (Cl. 2314.5)

This invention relates to a procedure for processing of materials containing the element of atomic number 94, known as plutonium, for separating the plutonium from extraneous matter such as substances of the kind present in neutron irradiated uranium exemplified by uranium and especially. fission products, and the like radioactive contaminants. More particularly, this invention concerns a separatory and concentration procedure involving the use of potassium lanthanum sulfate for forming a carrier precipitate.

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 93 having a mass of 239 is referred to as 93 Reference 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 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 and 94 is equal to approximately 0.02 percent by weight of the uranium mass. At this concentration of these substances, the concentration of fission elements which must be removed is approximately the same percentage.

A number of processes have already been proposed for accomplishing the separation and concentration of Pu involving the formation of'carrier precipitates. Certain of these processes are generically known as the bismuth phosphate type process and the wet fluoride type of process. These processes are the invention of others and the details of the processes are described in copending applications as for example App. Ser. No. 519,714, now Patent No. 2,785,951, issued March 19, 1957, to be referred to hereinafter, which gives details relative to such processes.

In accordance with the present invention, it is proposed to provide a new and alternative crystalline type of carrier precipitate and method of formation, which will carry Pu substantially quantitatively from the acidic solutions of the type usually encountered in plant operations. That is, the carrier precipitate of the present invention functions in a manner similar to those carrier precipitates described in App. Ser. No. 519,714, now Patent No. 2,785,951, issued March 19, 1957, but possesses certain advantages thereover.

The meaning of the terms carrier precipitate, byproduct precipitate, product precipitate, extraction and decontamination and other similar terms will be apparent as the description proceeds.

This invention has for one object to provide improvements in a method for the separation and recovery of Pu.

Another object is to provide a method for separating and recovering Pu wherein a carrier precipitate different from those heretofore used is employed.

Still another object "is to provide a process for forming a new type of carrier precipitate.

Another object is to provide a method for forming a type of carrier precipitate which exhibits improved characteristics such as crystalline structure and faster settling rate.

Another object is to provide a method of forming a crystalline type of carrier precipitate which method may be coupled with existing processes.

Still another object is to provide a method of forming a carrier precipitate utilizing potassium lanthanum sulfate.

Still further object is to provide methods for forming a crystalline type of lanthanum carrier precipitate.

Still further object is to provide a type of process which may be carried out in existing equipment.

Other objects will appear hereinafter.

I have found that a crystalline type of carrier precipitate may be formed utilizing potassium lanthanum sulfate which will carry plutonium substantially quantitatively. Although potassium lanthanum sulfate is nitric acid soluble and while the solutions containing plutonium frequently encountered industrially are nitricacid solutions, by means of the process of the present invention, a satisfactory crystalline precipitate may be obtained. In general, the formation of the crystalline type of precipitate of the present invention would be carried out upon solutions containing Pu which have reached the concentration stage. That is, the solution processed by the present invention would preferably have been subjected to preliminary treatments such as extraction and decontamination for eliminating extraneous matter such as fission products and other components such as uranium. The use of the bismuth phosphate treatment is exemplary of such preliminary type of treatment.

An illustration of certain types of carrier precipitates and conditions of formation are described in App. Ser. No. '5 19,714 filed January 26, 1944, Thompson and Seaborg, now Patent No. 2,785,951, issued March 19, 1957,

and reference is made to that application for further disclosure, details thereof being omitted from the present state or states referred to herein as Pu in which the element is characterized by forming insoluble phosphates and fluorides and a higher oxidation state or states referred to as Pu in which the element forms soluble phosphates and fluorides. Plutonium in the lower oxidation state has a valence not greater than +4 and in the higher oxidation state has a valence greater than +4.

While processes of the above type such as the aforementioned wet fluoride type of process operates satisfactorily and carries large amounts of Pu these fluoride precipitates may be of a fine structure and gelatinous nature. Their rate of settling may be relatively slow compared with the carrier precipitate of the present inven tion. Consequently, by the present invention it is proposed to provide an alternative type of carrier precipitate which, for example, maybe used in place of the standard lanthanum fluoride type precipitate. vention it is also proposed to provide a crystalline type of precipitate which exhibits better settling properties and other improved characteristics.

The solutions containing Pu which may be treated by my invention may be the same type of solutions as heretofore treated. As indicated one common type of solution containing Pu subject to separation and recovery procedure are the solutions initially processed by a bismuth phosphate type of treatment. This type of solution and its treatment are described in detail in App. Ser. No. 519,714 aforementioned, now Patent No. 2,785,951, issued March 19, 1957. Such solutions comprise anitric acid containing liquid having a content of Pu therein. The nitric acid solution may also contain other materials such as contents of phosphoric acid and bismuth ions.

The solution may also containcertain extraneous matter I such as radioactive materials which the subsequently applied carrier precipitation treatments may eliminate-or reduce or carry the Pu away from. As referred to above, by means of a carrier precipitation treatment applied to the solution containing the Pu in the reduced condition, the resultant precipitate (product precipitate) carriesthe Pu away from extraneous matter. By means of a precipitation applied to the solution having the Pu in the oxidized condition (by-product precipitate) extraneous matter is carried away by the precipitate leaving the oxidized Pu in the supernatant liquid from this precipitation.

In either instance, by means of my invention, a precipitate may be formed which settles more rapidly and otherwise has better characteristics.

While various types of solutions containing Pu may be treated, the solution described herein as treated by the present invention are similar to these described in App. Ser. No. 519,714 aforementioned, now Patent No. 2,785,951, issued March 19, 1957. That is, the solutions have been subjected to a bismuth phosphate type of extraction and decontamination of one or more cycles until the solution is ready for concentration. Such solutions comprise dilute nitric acid containing the Pu. In general in applying my invention the solution would be processed by reducing the Pu therein to the reduced state. Then there is incorporated a content of fluoride ions as by introducing hydrogen fluoride into the solution. A content of a potassium lanthanum sulfate (crystalline) in accordance with the present invention, is incorporated preferably in the form of a slurry. A crystalline type precipitate without apparently dissolving carries down the plutonium substantially quantitatively. This carrier precipitate may comprise a fluoride derivative or other derivative of the potassium lanthanum sulfate as will be described in further detail hereinafter as some reaction may have taken place. However, the exact composition of the crystalline carrier precipitate has not been identified, hence it is not desired to be restricted to any particular composition or theory of operation.

In order to provide a further understanding of the operation of my invention and work carried out for study- By the present in- 4 ing the formation of the new crystalline type of precipitate, reference is made to the observations and examples which follow.

A solution of the .type described above, namely a solution resulting from prior bismuth phosphate treatment, 1 N in HNO was made 0.1 N in KNO A solution containing about 225 mg./l. La was added. Thereafter HF was added slowly. A LaF precipitate was obtained (relatively slow settling and gelatinous). This experiment was an attempt to obtain a potassium lanthanum fluoride typeof precipitate which would have different characteristics as regards settling than ordinary LaF In thenext experiment a solution of the type above described was made 0.5 M in K SO by adding a solution of K 50 containing 87 grams thereof per liter. Thereafter a solution containing La+ in a concentration of 225 mg./l. was added. The resultant liquid was digested. Apparently a potassium lanthanum sulfate type precipitate did not precipitate. Potassium lanthanum sulfate is soluble in l N HNO A content of HF was then added. The usual type ofLaF precipitate was obtained.

From a further group of experiments it was observed that if lanthanum nitrate or lanthanum ammonium nitrate were treated with K to precipitate a potassium lanthanum sufate and HF added to the slurry, a crystalline LaF KLaF. or LaF coated with potassium lanthanum sulfate apparently may be formed. The addition of 1 N HNO did not dissolve this precipitate. If a slurry of this potassium lanthanum sulfate material is added to a 0.2 N HF1 N HNOg solution containing Pu, dissolution does not take place. The crystalline form apparently remains the same. The potassium lanthanum sulfate apparently is soluble in water but highly insoluble in excess K 50 The supernatant from a precipitation of potassium lanthanum sulfate was void of La+ as shown by adding HP to the supernatant. In the light of the preceding observations, runs were tried for preparing potassium lanthanum sulfate and employing this addition for caus ing carrying of plutonium. Reference is now made to examples of this procedure.

Example I About 263 mg..of pure La(NO containing 112.5

mg. La per liter were treated with 10 ml. of saturated K 80 solution. A potassium lanthanum sulfate precipitated in small crystals which upon agitation grew rapidly giving a rapid settling slurry.

A starting solution comprising 500 ml. of a usual type of solution from preliminary bismuthphosphate treatment was made 0.4 N in HF. A slurry of potassium high a concentration of Lo which forms a gelatinous LaF The potassium lanthanum sulfate preferably is not added to the oxalic acid reduced solution, namely the solution as obtained from prior bismuth phosphate treatment, directly in the absence of HF as dissolution of the addition may take place.

Example ll About 500 ml. of a plant, oxalic acid (11 C 0 reduced, solution was made 0.4 N in HF. That is, a solution which had been given preliminary bismuth phosphate treatment was processed by the present invention. A crystalline carrier precipitate was formed from lanthanum ammonium nitrate dissolved in 0.2 ml. H 0 and ml. of 0.2 M K 80 added thereto. The potassium lanthanum sulfate precipitate obtained was agitated for /2 hour then added to the 500 ml. of solution aforementioned over a period of 30 minutes. The resultant liquid containing precipitate was digested 1% hours at room temperature. The precipitate carrying Pu was separated by centrifuging. The crystalline precipitate easily separated. The precipitate was washed with dilute HNOg-HF. A rough analysis of waste volume of 565 ml. by direct evaporation'showed that 98.4% Pu was carried by the carrier precipitate. Another waste analysis by a different method (precipitation with LaF carrier) showed that 98.7% of the Pu had been carried.

In order to check the recovery further it was thought that since it had been somewhat diflicult to pour off the supernatant cleanly that perhaps some of the product lost in the waste may have been due to solid matter carried over. Samples of waste supernatants were analyzed for Pu (product) content. From this analysis indicating only 0.85% Pu in the waste, it appears that the plutonium was carried to the extent of 99.1%. Therefore it appeared that the crystalline carrier of the present invention carried the Pu to the extent of at least 99%.

While it is to be noted that potassium lanthanum sulfate is soluble in nitric acid, as described above by incorporating a source of ions such as fluoride ions into the nitric acid solutions containing Pu processed, the dissolution of the precipitate is prevented and a crystalline carrier is present which will substantially quantitatively carry down the Pu. The composition of the carrier addition of the present invention may be designated by the formula La (SO .3K SO However larger amounts of K 80 may be present. As indicated above excess K 80 tends to render the precipitate insoluble and may be present.

3 It may be that some of the potassium lanthanum sulfate reacts to form a potassium lanthanum fluoride. Some lanthanum fluoride may be formed which becomes coated with the crystalline carrier of the present invention or other reactions or rearrangements may take place. It is suflicient to indicate that a crystalline type of precipitate is present which carries Pu substantially quantitatively and that this carrier is diflerent from other carriers in having a faster settling rate, for example. By applying the process to an oxidized solution, namely a solution containing oxidized Pu, a crystalline by-product precipitate is formed which carries extraneous matter such as fission products.

The process of the present invention may be applied to solutions containing Pu from tracer amounts to several hundred grams thereof per liter of solution. The concentration'of fluoride ion as for example hydrogen fluoride may be from about .2 to 1 M. The source of fluoride ion may alsobe accomplished by adding potassium and sodium fluoride or other similar soluble fluoride materials. The exact method employed for reducing Pu or other similar details constitute known practice and is not a limitation on my invention.- These steps and other similar operations have been set forth merely as guides to preferred practice and are not a limitation on the present invention.

It is to be understood that all matter contained in the above description and examples shall be interpreted as ilustrative-and not limitative of the scope of this invention, and it is intended to claim the present invention as broadly as possible in view of the prior art.

I claim:

1. A process for separating plutonium values in an oxidization state not greater than +4 from an aqueous fluoride solution, which comprises introducing a crystalline potassium-lanthanum sulfate into the solution whereby a crystalline lanthanum fluoride carrier precipitate is formed, and separating said lanthanum fluoride carrier precipitate together with associated plutonium values from the solution.

2. A process for separating plutonium values in an oxidization state not greater than +4 from fluoride-soluble radioactive fission products, which comprises forming an aqueous acidic solution containing plutonium values in said valence state, said fission products, and

fluoride ions, introducing a crystalline alkali metal lanthanum sulfate into said solution whereby a crystalline lanthanum fluoride carrier precipitate is formed and separating said lanthanum fluoride carrier precipitate with the associated plutonium values, leaving the fluoride-soluble fission products in solution.

.3. The process of claim 2 wherein the alkali metal component of the alkali metal lanthanum sulfate is potassium.

4. The process of claim 2 wherein the alkali metal component of the alkali metal lanthanum sulfate is sodium.

5. The process of separating plutonium values in an oxidization state not greater than +4, from fluoridesoluble radioactive fission products, which comprises forming an approximately 1 N HNO solution containing plutonium values in said valence state, fluoride-soluble fission products, and fluoride ions in a concentration of approximately 0.2 to l M, introducing a crystalline potassium-lanthanum sulfate precipitate into said solution whereby a crystalline lanthanum fluoride carrier precipitate is formed and separating said lanthanum fluoride carrier precipitate together with associated plutonium values from the solution, leaving the fluoride-soluble fission products in solution.

References Cited in the file of this patent UNITED STATES PATENTS Thompson et al. Mar. 19, 1957 OTHER REFERENCES 7 declassified Nov. 22, 1957, which reports as bibliographic reference 37 ABC Document CN-522, to Stein, pp. 11, 12, Mar. 15, 1943.

Claims (1)

1. A PROCESS FOR SEPARATING PLUTONIUM VALUES IN AN OXIDATION STATE NOT GREATER THAN +4 AND FROM AN AQUEOUS FLUORIDE SOLUTION, WHICH COMPRISES INTRODUCING A CRYSTALLINE POTASSIUM-LANTHANUM SULFATE INTO THE SOLUTION WHEREBY A CRYSTALLINE LANTHANUM FLUORIDE CARRIER PRECIPITATE IS FORMED, AND SEPARATING SAID LANTHANUM FLUORIDE CARRIER PRECIPITATE TOGETHER WITH ASSOCIATED PLUTONIUM VALUES FROM THE SOLUTION.