US2868619A

US2868619A - Process for the recovery of plutonium

Info

Publication number: US2868619A
Application number: US558451A
Authority: US
Inventors: David M Ritter
Original assignee: Individual
Current assignee: Individual
Priority date: 1944-10-12
Filing date: 1944-10-12
Publication date: 1959-01-13
Anticipated expiration: 1976-01-13

Description

PROCESS FOR THE RECGVERY F PLUTONIUM David M. Ritter, Oak Ridge, Tenn, assignor to the United States of America as represented by the United States Atomic Energy Commission Application Gctober 12, 1944, Serial No; 558,451

8 Claims. (or. 23-145 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 a special combination of steps.

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 and the term plutonium value 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 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 separation and concentration of Pu. Certain of these processes are generically known as the bismuth phosphate type process and the wet fluoride" rQQ type of process. These processes are the invention of others and the details of the processes are described in copending application, as for example app. Ser. No. 519,714, now U. S. Patent No. 2,785,951, to be referred to hereinafter, which gives complete details relative to the bismuth phosphate process. Consequently, all the details of operation of the aforementioned processes are not described herein.

While the aforementioned processes have been used successfully, it is proposed by the present invention to describe certain alternative procedure for obtaining additional advantages. That is, by the present invention utilizing many of the same steps which have been previously worked out, but in a different combination and arrangement, advantages of improved decontamination may be obtained. In addition to carrying out the steps in a different order in accordance with the present invention, certain new'steps may be interposed for facilitating the rearrangement of process order.

The term decontamination as used herein refers in particular to the separation or removal of radioactive material from the Pu, or materials containing the Pu. The meaning of the term carrier precipitate and other similar terms will be apparent as the description proceeds.

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

Another object is to provide a method of separating plutonium by procedure wherein a different combination of steps than has heretofore been used is employed.

A still further object is to provide a process for separating Pu wherein elimination of extraneous materials such as fission products may be better accomplished.

Still another object is to provide a novel type of process for separating the Pu which lends itself to the use of and to coupling with steps already known or precticed.

Another object is to provide a novel type of separatory process for the recovery of Pu which may employ many of the materials used in existing processes and which may be carried out in existing equipment without change, or with a minimum of equipment change.

Still another object is to provide a process for the recovery of Pu which may be applied to solutions containing Pu either in relatively small or large amounts.

Another object is to provide a separatory and concentration process for Pu containing materials involving the use of both lanthanum fluoride and bismuth phosphate type carrier.

Other objects will appear hereinafter.

For a further understanding of the invention, reference will later be made to the attached drawing forming a part of the present application. In this drawing, a diagrammatic representation of one embodiment of the invention is given in the form of a flow sheet.

I have found that Pu in admixture with various extraneous material may be separated and concentrated with-a high degree of decontamination by the use of a series of steps involving following a lanthanum fluoride treatment with a bismuth phosphate treatment, rather than in the converse order as has heretofore been practiced. That is, the present process involves the use of known steps such as the use of bismuth phosphate precipitation, but it has been found that by rearranging the order of the steps and by making certain additions which are described in detail hereinafter, that a process is obtained wherein greater decontamination may be accomplished while at the same time there may also be obtained product separation, reduction in carrier bulk, and other of the advantages obtained in certain regular processes.

The bismuth phosphate type of extraction is set forth in application Serial Number 519,714, filed January 26,

1944, Thompson and Seaborg, now U. S. Patent No. 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 wherenecessary to an understanding of the present invention. As set forth in said application, it has been discovered 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 a higher oxidation state or states referred to as Pu in which the element forms soluble phosphates and fluorides.

It has been observed that the aforementioned processes, namely the bismuth phosphate process and the wet fluoride process in the order named may not give as high values for decontamination as might be desired. 1 have discovered from fission product analysis that in general fission products may hold back during the bismuth phos phate precipitation to a degree greater than they carr".

That is, some ionic species dissolved in the presence of a precipitate tend to partition between the precipitate and the solution. That portion of the ionic species which remains in solution is said to hold-back. The portion that follows the precipitate is said to carry.

During the wet fluoride treatment, namely, during a lanthanum fluoride precipitation step, the carrying of fission product activity predominates over the hold-back.

Therefore, it is apparent that if the carrying of product and the hold-back of fission materials which occurs during the reduced product phases of the bismuth phosphate process could be combined with the lanthanum fluoride fission product carrying in the oxidized product phases, a

substantial increase in decontamination per cycle may be accomplished.

l have found that such a transfer may be facilitated by interposing a treatment described herein as complexing whereby a bismuth phosphate cycle may be caused to readily follow a fluoride treatment. By this complexing treatment, fluoride ions which might be expected tointerfere in the bismuth phosphate cycle are successfully solubilized and prevented from causing any difficulty. The term complexing as used herein embraces the solubilizing or other dis osal of fluorine ions in a manner to prevent such ions fro-m hampering succeeding steps.

in general, the operation of my processes for treating a nitric acid solution containing Pu is as follows: The solution is treated by a standard procedure for separating Pu away from the U, uranyl nitrate hexahydrate referred to in the art as UNI-l and other bulk components. This may be accomplished in any conventional manner. The

material containing Pu is redissolved and oxidized so that a standard by-product' precipitation treatment may be applied. The solution resulting from the by-product precipitation containing Puin an oxidized state is then subjected to a fluoride lay-product precipitation wherein substantial decontamination is accomplished. The liquid from this step still containing Pu in the oxidized state is then treated with a suitable complexing agent or equivalent material to prevent interference in further steps by any fluorine ions present. The complexed oxidized solution is then reduced and subjected to the standard bismuth phosphate product precipitation which carries down the Pu freed of fission products or other contarninating materials.

in carrying out of the process, the various precipitates referred to may be separated by filtration, centrifugation or other conventional manner as usual in processes of this type.

A further understanding of the general application of my process may be had by reference to the attached flow sheet. It is there indicated that the product containing solution is subjected at l to a standard extraction step for removing the product Pu away from a substantial amount of the uranium and related bulky material. The

carrier containing Pu, freed from bulky components is then redissolved at 2 and oxidized. A holding oxidant may be present as indicated at 3. A standard fluoride byproduct precipitation is applied to the oxidized solution at 4 thereby eliminating troublesome fission product cantamination remaining with the Pu at this step. Some contamination has been removed in preceding steps.

As indicated above, it has been discovered that by lanthanum fluoride precipitation the carrying of fission products and the like is supeiror to the hold-back thereof, hence, substantial decontamination is accomplished. However, by this decontamination step, the resultant decontaminated solution containing Pu may contain a corn tent of a fluorine ion which may hamper recovery of Pu from the solution in further processing. it has been found that in accordance with the present invention that any such ions present may be easily complexed by the addition of suitable reagents at 5. Thereafter, the complexed solution is reduced at 6. The reduced solut n then has applied thereto a standard bismuth phosphate product precipitation step in which Pu is quantitatively separated by the carrier from the reduced solution. This separation is in no way materially hampered by the prior addition of fluoride materials. This Pu precipitation step is indicated at 7. The precipitate carrying Pu is separated by centrifuging or filtering and may be further treated in any desired manner at step 8 to isolate Pu or convert the Pu to other derivatives.

The complexing step of the present invention referred to under 5 in the flow sheet may comprise the incorporation in the solution of any reagent that will deter or prevent the fluorine ion from hampering the process in subsequent steps. Among the various materials which may be used for this purpose mention may be made of various boron compounds. However, other reagents which will complex fluoride ions or otherwise similarly function may be employed.

The new combination of steps including the aforementioned feature of complexing fluoride ions are set forth below in further detail. As already indicated, certainof the steps are individually old, or are not per se regarded as the invention of the herein-named inventor. However, the combination of steps set forth below and particularly steps (4) is believed to constitute novel procedure. Referring now to a detailed example, the steps are as follows:

(1) The material from which Pu is to be recovered is subjected to a standard bismuth phosphate extraction procedure. Broadly this procedure may include dissolving the material containing Pu in a solvent such as nitric acid. The resultant solution is reduced with formic acid, ferrous ion, uranous ion or other reducing agent which have been found to be satisfactory. The resultant reduced solution, in diluted condition, is treated with sulfuric acid or other source of sulfate ion for complexing the UO A source of bismuth ions, such as BiONO is added, and a bismuth phosphate product precipitate obtained by the addition of phosphoric acid.

(2) The resultant precipitate, carrying Pu, obtained from step (1) is redissolved in a solvent such as nitric acid and subjected to sodium bismuthate oxidation. Preferably the oxidation treatment isaccompanied by the use of a di:hromate or comparable soluble reagent as a holding oxidant. As indicated, this feature and certain others form the subject matter of related patent applications. After oxidation is complete, a bismuth phosphate by-prcduct precipitation is accomplished by adding phosphoric acid. The by-product is centrifuged out leaving the centrifugate liquid containing Pu in the (0) condition.

(3) The solution from the preceding step, containing Pu in the (o) state, is subjected to a wet fluoride treatment such as a lanthanum fluoride by-product precipitation. This is accomplished by adding a source of lanthanum ions to the oxidized solution, as for example,

adding -suflicient lanthanum nitrate to give a concentration of about .225 gram of lanthanum per liter. Then a source of fluoride ions such as H F is added to make the solution about 1 M in H F- Preferably a content of a dichromate or similar holding oxidant is present during this step. A LaF by-product precipitate is thereby thrown down which may be centrifuged out leaving centrifugate containing Pu in the (o) state. Troublesome radioactive materials are carried out by the lanthanum fluoride precipitate. In other words, good decontamination is obtained.

(4) Referring now to the steps pertaining in particular to special aspects of my invention, the solution from the preceding step containing Pu in the state, and any residual H F is treated with at least one of the boron compounds H BO (boric acid), NaBO (sodium metaborate) or Na B OlOH O (borax). Other comparable materials may be used for complexing this residual H 1 Addition of a boron material such as H BO to 1 N H F 1 N HNO prevents precipitation of BiF and BiOF upon addition of a source of bismuth ions as by adding bismuth nitrate. No precipitation occurs under these circumstances even upon digestion at 75 C. The addition of H PO, to such a solution, however, results in the immediate formation of BiPO (5) Because of the complexing of the fluoride ions in the preceding step, the solution containing Pu in the (0) state may now be subjected to the standard bismuth phosphate product precipitation. This is accomplished by reducing the solution with ferrous ion, uranous ion, oxalic acid, hydrogen peroxide or other reducing treatment and obtaining a precipitation of product Pu by the addition of BiONO H PO etc. as described under step (1.).

(6) The bismuth phosphate product precipitate, carrying Pu may then be treated further, if desired, for isolating or reacting the Pu to form derivatives. One convenient procedure comprises passing fluorine gas through the precipitate containing the Pu. In this manner the Pu may be separated as a volatile compound and condensed.

The introduction of boron into the process is not objectionable on the basis that boron might appear in the final product. If the dry fluoride process (passing F in contact with the precipitate) or similar process is employed for further treatment, any boron or equivalent complexing residues will be removed. For example, boron will be removed as BF (B. P. 101 C.) which would in many instances be the most volatile component of the final mixture. As indicated, other purification procedures will operate equally Well.

One of the advantages of the process above described is that more efiicient product carrying and decontamination may be obtained by the combination of steps above indicated. At the same time reduction of volumes and other advantages obtainable in standard processes are still obtainable. The process of the present invention is further advantageous in that the same reagents, procedure and apparatus that are used in existing processes may, in a number of steps, be used in the present process.

It will be noted in the present invention, however, that although certain known steps are employed that their order of use is dififerent. That is, the fluoride precipitation step precedes the phosphate precipitation in the combination of the present invention. Also a complexing or similar step as already described may be interposed in the present invention.

According to the best evidence available, the oxidation :state of plutonium secured in solution by the action of the oxidizing agents referred to herein and in cited copending application Ser. No. 519,714, new U. S. Patent No. 2,785,951, is greater than four, and the oxidation -:state secured in solution by the action of the reducing ,agents referred to herein and therein is no greater than ;.four. It should therefore be understood that, as used 6 herein, Pu and Pu in the (0) condition (or state) are both synonymous with plutonium in an oxidation state greater than four and Pu and Pu in the (r) condition (or state) are both synonymous with plutonium in an oxidation state no greater than four.

It is to be understood that all matters contained in the above description and examples shall be interpreted as illustrative 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. In processes for the decontamination and recovery of plutonium values from an aqueous, acidic solution containing the same together with contaminant, comprising selective carrier precipitation of at least a portion of said contaminant from said solution away from said plutonium values by means of precipitation therein of lanthanum fluoride while maintaining the plutonium in an oxidation state greater than four, and removal of the resulting contaminant-containing carrier precipitate from the remaining saidsolution, the improvement method which comprises combining with said contaminant carrier precipitation and removal, the step of subsequently incorporating in said remaining solution a boron compound chosen from the group consisting of boric acid, sodium metaborate, and borax, to form a soluble complex with fluoride ion consequently present in said remaining solution, and thereafter precipitating said plutonium values from the resultingly-complexed solution by providing therein a sufficient concentration of bismuth ions and phosphate ions to precipitate bismuth phosphate while maintaining the plutonium in an oxidation state no greater than four.

2. In processes for the decontamination and recovery of plutonium values from an aqueous, acidic solution containing the same together with contaminating uranium fission products, the improvement method which comprises, in combination, precipitating at least a portion of said uranium fission products from said solution away from said plutonium values by providing therein a sufiicient concentration of lanthanum ions and fluoride ions to precipitate lanthanum fluoride as a carrier while maintaining the plutonium in an oxidation state greater than four, then removing the resulting fission-product-containing carrier precipitate from the remaining said solution, subsequently incorporating in said remaining solution a boron compound chosen from the group consisting of boric acid, sodium metaborate, and borax, to form a soluble complex with fluoride ion consequently present in said remaining solution, and thereafter precipitating said plutonium values from the resultingly-complexed solution I by providing therein a sufi'icient concentration of bismuth ions and phosphate ions to precipitate bismuth phosphate while maintaining the plutonium in an oxidation state no greater than four.

3. The method of claim 2 wherein said boron compound is specifically boric acid.

4. The method of claim 2 wherein said boron compound is specifically sodium metaborate.

5. The method of claim 2 wherein said boron compound is specifically borax.

6. In processes for the decontamination and recovery of plutonium values from an aqueous, acidic solution containing the same together with contaminating uranium fission products, the improvement method which comprises, in combination, precipitating at least a portion of said uranium fission products from said solution away from said plutonium values by providing therein a suflicient concentration of lanthanum ions .and fluoride ions to precipitate lanthanum fluoride while maintaining the plutonium in an oxidation state greater than four, then removing the resulting fission-product-containing carrier precipitate from the remaining said solution, subsequently incorporating in said remaining solution a boron compound chosen from the group consisting of boric acid,

,values from the resultingly-complexed solution by providing therein a sufiicient concentration of bismuth ions and phosphate ions to precipitate bismuth phosphate while maintaining the plutonium in an oxidation state no greater than four, then separating the resulting plutonium-containing carrier. precipitate from its supernatant solution,

and thereupon passing fluorine gas through said separated precipitate to remove by volatilization any said boron compound concomitantly carried from solution therewith.

7. In processes for the decontamination and recovery of plutonium values from an aqueous, acidic solution containing the same together with contaminating uranium fission products, comprising carrier precipitation of said plutonium values accompanied by a portion of said fission products from said solution by means of precipitation therein of bismuth phosphate while maintaining the plutoniurn in an oxidation state no greater than four, then removal of the resulting plutonium-containing carrier precipitate from its supernatant and dissolution thereof to form an aqueous, acidic solution of said plutonium,

thereafter selective carrier precipitation from the formed solution of at least a portion of said fission products consequently contained therein away from said plutonium values by means of precipitation therein of lanthanum fluoride while maintaining the plutonium in an oxidation state greater than four, and removal of the resulting fission-product-containing carrier precipitate from the remaining said formed solution, the improvement method which comprises combining with said bismuth phosphate and lanthanum fluoride carrier precipitation steps, the steps of subsequently incorporating in said remaining solution a boron compound chosen'from the group consisting of boric acid, sodium metaborate, and borax to form a soluble complex with fluoride ion consequently present in said remaining solution, and thereafter precipitating said plutonium values fromv the resultinglycomplexed solution by providing therein a sufiicient concentration of bismuth ions and phosphate ions to preipitate bismuth phosphate while maintaining the plutonium in an oxidation state no greater than four.

8. In processes for the decontamination and recovery of plutonium values from an aqueous, acidic solution containing the same together With contaminating uranium fission products, comprising carrier precipitation of said plutonium values accompanied by a portion of said fission products-fron1 said solution by means of precipitation therein of bismuth phosphate while maintaining the plutonium in an oxidation state no greater than four, then removal of the resulting plutonium-containing carrier precipitate from its supernatant and dissolution, thereof to form an aqueous acidic solution of said plutonium,

thereafter 'selective carrier precipitation from the formed solution ofat least a portion of said fission products consequently contained therein away from said plutonium values by means OfPIccipitation therein of lanthanum fluoride while maintaining the plutonium in an oxidation state greater than four, and removal of the resulting fission-product-containing carrier precipitate from the remaining said formed solution, the improvement method which comprises combining with said bismuth phosphate and lanthanum fluoride carrier precipitation steps, the steps of subsequently incorporating in said remaining solution a boron compound chosen from the group consisting of boric acid, sodium metaborate, and borax, to

form a soluble complex, with fluoride ion consequently present in said rem'aining solution, thereafter precipitating 'said plutonium values from the resultingly-complexed solution by providing therein a sufiicieut concentration ofbismuth ions and phosphate ions to precipitate bismuth,

phosphate while maintaining the plutonium in an oxidation state no greater than four, then separating the resulting plutonium-containing carrier precipitate frcmiits' supernatant, solution, and thereupon passing fluorine gas through said separated, precipitate to remove by volatilization any 'said'boron compound concomitantly carried from solution therewith. a

OTHER REFERENCES Harvey: Nucleonics, 'April 1948, pp. 3040, particularly pages 34, 35 V M eller: Inorganic and Theoretical Chemistry, vol. 5,

Claims

1. IN PROCESSES FOR THE DECONTAMINATION AND RECOVERY OF PLUTONIUM VALUES FROM AN AQUEOUS, ACIDIC SOLUTION CONTAINING THE SAME TOGETHER WITH CONTAMINANT, COMPRISING SELECTIVE CARRIER PRECIPITATION OF AT LEAST A PORTION OF SAID CONTAMINANT FROM SAID SOLUTION AWAY FROM SAID PLUTONIUM VALUES BY MEANS OF PRECIPITATION THEREIN OF LANTHANUM FLUORIDE WHILE MAINTAINING THE PLUTONIUM IN AN OXIDATION STATE GREATER THAN FOUR, AND REMOVAL OF THE RESULTING CONTAMINANT-CONTAINING CARRIER PRECIPITATE FROM THE REMAINING SAID SOLUTION, THE IMPROVEMENT METHOD WHICH COMPRISES COMBINING WITH SAID CONTAMINANT CARRIER PRECIPITATION AND REMOVAL, THE STEP OF SUBSEQUENTLY INCORPORATING IN SAID REMAINING SOLUTION A BORON COMPOUND CHOSEN FROM THE GROUP CONSISTING OF BORIC ACID, SODIUM METABORATE, AND BORAX, TO FORM A SOLUBLE COMPLEX WITH FLUORIDE ION CONSEQUENTLY PRESENT IN SAID REMAINING SOLUTION, AND THEREAFTER PRECIPITATING SAID PLUTONIUM VALUES FROM THE RESULTINGLY-COMPLEXED SOLUTION BY PROVIDING THEREIN A SUFFICIENT CONCENTRATION OF BISMUTH IONS AND PHOSPHATE IONS TO PRECIPITATE BISMUTH PHOSPHATE WHILE