US2990241A - Improvement in decontamination of aqueous acidic solutions containing plutonium and fission product values by providing cerous and/or mercuric ions therein prior to a bismuth phosphate carrier precipitation - Google Patents

Description

United States Patent Burt F. Paris and Harry K. Strassel, Richland, Wash., as-

slgnors to the United States of America as represented by the United States Atomic Energy Commission I No Drawing. Filed Mar. 31, 1945, Ser. No. (586,038

5 Claims. (Cl. 23-145) This invention relates to a procedure for the 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 supplemental additions, such as mercury and cerium agents, whereby improved decontamination and Pu recovery may be obtained.

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 obtainedfrom 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, par- 1 ticularly 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 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 nu- .cleus 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 he removed is approximately the same percentage.

A number of processes have already been proposed for ice accomplishing the separation and concentration of Pu. 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 othersand the details of the processes are described in copending applications, as for example application Ser. No. 519,714, filed January 26, '1944, now US. Patent 2,785,951, issued March 19, 1957, to be referred to here inafter, which gives details relative to such processes.

It is sufiicient to indicate at this point that by means of such processes, of which there are a number, applied in one or several cycles dependent upon the concentration of contaminants and similar considerations that thecontamination may be substantially reduced and the plutonium recovered. That is, as described above the desired plutonium to be recovered is in the presence of substantial quantities of fission products evidenced by beta and gamma activities, and the plutonium must be isolated from at least a portion of the activity. In other words, the activity must be eliminated either completely or at least to a. certain tolerance level so that the Pu may be handled or further used without danger from radioactivity due to contaminants. The bismuth phosphate method of by-product and product precipitation is generally illustrative of processes which may be applied to separate and recover plutonium.

While existing processes, includingthe bismuth phos phate type of process, are quite efficient and successful, it may be necessary to repeat a process a numberof times in order to obtain a Pu in which the activity .has been reduced to the tolerable amount. This may be accomplished by repeated application of product and by-product cycles as will be referred to in further detail hereinafter. In some instances, the feed materials to be processed may be higher in both plutonium content and much higher in gamma activity than materials usually processed. Consequently, there is a need for supplemental procedures in order to more quickly and thoroughly redu'ce oreliminate activities.

We have found that the aforementioned processes which are referred to herein as existing, standard or conventional processes, exemplified by the bismuth phosphate process, may have included therein certain supplemental agents and operations for materially improving decontamination. This improvement may be indicated by obtaining a favorable decontamination factor in fewer cycles, or by obtaining a higher decontamination factor, namely, more complete decontamination than formerly possible by the particular process in the same number of steps.

The meaning of the terms bismuth phosphate type of process, product and by-product precipitate, decontamination, and similar terms are apparent to some extent from the preceding description and will be further apparent as the description proceeds.

The invention has for one object, to provide improvements in a method for the separation and recovery of plutonium.

Another object is to provide a method of separating plutonium by procedure wherein certain agents and steps supplemental to those ,hereintofore used are employed.

Another object is to provide improved procedure for eliminating, irradicating or reducing gamma contamination.

Still another object is to provide improved procedures particularly useful in the bismuth phosphate type of proc- 6S5.

Another object is to provide improved procedures in forming both by-product and product carrier precipitates.

Still another object is to provide a procedure for improving decontamination.

Still a further object is to provide steps and reagents which lend themselves to combination with steps already known or practiced.

Another object is to provide a type of process which may employ materials used in existing processes, but with more effective decontamination, and which may be carried out in existing equipment without change, or with a minimum of equipment change.

Still another object is to improve existing procedures by reducing the number of cycles required for decontamination or obtaining higher decontamination by the same number of cycles.

Another object is to provide a process susceptible of treating sources of Pu containing concentrations thereof and of activities higher than heretofore usually encountered.

Other objects will appear hereinafter.

We have found that Pu in admixture with various extraneous material may be separated and concentrated by the use of the series of steps involving the formation of certain carrier precipitate type of treatments. These treatments may be similar to existing practice in many features as respects some of the reagents, temperatures and similar aspects. However, we have found that the function of such carrier precipitates of carrying contamination away from Pu and carrying Pu away from contamination may be rendered considerably better by the inclusion of certain reagents and steps as set forth in detail herein. By the formation of the carrier precipitate under the conditions of the present invention, not only are the advantages, such as good separation of Pu previously obtained in the processes, still obtainable, but additional advantages such as improved gamma decontamination are obtained.

As referred to above, an illustration of the types of carrier precipitates involved are described in application Ser. No. 519,714, filed January 26, 1944, Thompson and Seaborg, now US. Patent 2,785,951, issued March 19, 1957, and reference is made to that application for further disclosure, details thereof being omitted from the present disclosure except where necessary 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 other insoluble compounds and a higher oxidation state or states referred to as Pu in which the element forms soluble phosphates and other soluble compounds. As further set forth in said application Ser. No. 519,714, the soluble plutonium phosphate in which plutonium is present in its higher oxidation state is referred to as plutonyl phosphate, (PuO (PO Pu therein having a valence of six, and the valence of plutonium in its phosphate insoluble state is not greater than four.

That is, by having the metal as Pu" a product precipitation or extraction may be accomplished in which a carrier precipitate brings down the Pu" leaving behind in solution a substantial portion of fission products and other sources of contamination. However, some of the activity may be carried along with the precipitate or otherwise may be occluded with the Pu" Certain contaminants such as columbium, zirconium, and lanthanum activities may be quite difiicult to separate from -Pu.

The carrier precipitate containing Pu" after suitable dissolution, is oxidized so that Pu is obtained which remains in solution and by means of a by-product carrier precipitate, such as a bismuth phosphate by-product precipitate, under the oxidizing conditions carriesdown fission products leaving the Pu in solution. However, in this step certain by-pro'duct precipitate fines may not separate out or otherwise some contamination remains so that the solution remaining which contains the Pu may still be contaminated with some fission products although in much smaller amounts. It should be kept in mind that since the Pu is in an environment of radioactivity which may give many million of counts a minute per milligram of fission product present, that the clean-cut separation of the Pu presents considerable difficulty.

Illustrative details respecting bismuth phosphate byproduct and product precipitation will be apparent from the following: Neutron irradiated uranium is dissolved in a suitable quantity of 60-70% nitric acid. This gives a solution containing Pu such as may be referred to as a UNH (uranyl nitrate hexahydrate) solution. The solution is subjected to treatment with an agent such as sodium nitrite, formic acid, H 0 oxalic acid or the like 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 UNH is adjusted to 20% and H fluo silicic acid or the like 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% UNH; phosphoric acid is also added to make the solution between .3.8 M therein, and a precipitate comprising BiPO which carries the Pu comes down and is separated from the solution by filtration or centrifugation. This illustrates a product precipitation. 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 Cr O Various other oxidizing agents such as sodium bismuthate may also be used. 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 l 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 to about C. whereupon BiPO precipitates carrying fission products but not Pu The precipitate may be removed by filtration or centrifugation and discarded. This illustrates a by-product precipitation. 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, or by oxalic acid, Fe+ or similar reduction, and the cycle is suitably repeated. 7

The exact manner of obtaining the bismuth phosphate product and by-product precipitates is not a limitation on the present invention. The product precipitation under (r) conditions, as indicated, may be accomplished by using any of a number of reducing agents of which sulfur dioxide, hydrogen peroxide, oxalic acid, sodium nitrite, ferrous iron, and the like are mentioned for illustration. Likewise the by-product precipitation under (0) conditions, as indicated, may be accomplished by a number of oxidizing agents for obtaining the (0) state. Dichromates, sodium bismuthate, permanganate and the like are mentioned for illustration. Various amounts and sources of bismuth or phosphate ions may be used. Details respecting these aspects are described in other copending applications and form no integral part of the present invention. Consequently, in the description which follows reference will merely be made to a bismuth phosphate product or by-product carrier, or similar terms used, it being apparent that such carrier may be'formed in any suitable manner.

The solutions containing Pu which may be treated by our invention may be the same type of solutions as heretofore treated, or may be solutions in which the concentrations of components are much higher. As apparent from the foregoing, one common type of solution containing Pu subject to separation and recovery procedures are the solutions processible by a bismuth phosphate type. of

treatment. As indicated, such solutions comprise a nitric acid containing liquid having a content of Pu therein. The nitric acid solution may also contain other'materials such as a content of various other acids. The solution may also contain certain extraneous matter such as radioactive materials which the procedure of the present invention eliminates or reduces.

We have found that by means of certain supplemental additions, such as a cerium addition applied to the solution containing the Pu namely in the reduced condition, the carrier (product precipitate) better carries the Pu away from extraneous matter. By means of certain supplemental additions, such as a mercury addition, applied to the solution having the desired element Pu in a state of higher oxidation (by-product precipitate) extraneous matter is carried away by the precipitate leaving the Pu in the supernatant liquid from this precipitation thereby decontaminated to an improved extent.

Broadly, therefore, our invention may be applied to various solutions and particularly solutions containing high activities. The solutions to be processed by our invention have added thereto such reagents'as may be required, dependent upon the particular composition and condition of the solution, to obtain the customary carrier precipitate formation at that particular phase of the process. Before, simultaneously therewith, or thereafter as will be further apparent from the examples, there is applied thereto the procedure of the present invention for causing improved decontamination, particularly the elimination of gamma activity.

In general, the operation of our process as respects its use in conjunction with bismuth phosphate precipitates is as follows: We have found that the addition of a source of mercury ions such as mecuric nitrate, preferably added before a bismuth phosphate by-product precipitation step, improves decontamination. That is, the addition of Hg++ increases the elimination of gamma activity in both the by-product and the product precipitation step. There would also be added before the next product precipitation step a source of cerium ions. By using a combination of agents such as mercury in the byproduct precipitation and cerium in the product precipitation step, for example, a four-fold increase in gamma decontamination may be efiected. Also the plutonium separation in the product precipitation step is better. Approximately, 1-5 grams of Hg++ per liter of solution treated and 1-5 grams of Ce+++ are illustrative of quantities of these supplemental additions which may be used under most conditions.

A still further understanding of our invention will be had by reference to the following detailed examples. In these examples the solutions containing Pu which was treated were, in general, the usual type encountered industrially, excepting that the gamma activity was in many instances higher than usual. That is, the solutions were obtained from standard processing and were approximately 1 normal in nitric acid. The solution also may contain residual oxidizing of reducing agents, various inorganic acid and the like ingredients in addition to extraneous material, such as high activities.

EXAMPLE I In this example a nitric acid solution of the abovedescribed type was treated and agents of the present invention were added in two places. One hundred cubic centimeters of starting solution (extraction product dissolved in N HNO 2.5 g./l. Bi as BiPO was oxidized and 5 g./l. Hg, as mercuric nitrate, was added before a BiPO by-product precipitation was made. The byproduct filtrate was reduced and 2 g./l. tartaric acid added, and the BiPO final product precipitation made.

The results of the example indicated improvement in decontamination. Almost twice as much gamma activity was taken out by the by-product precipitation, only 12.0% of the remaining gammas were carried by the final product precipitation, and the gammas remaining at the V 6 r 7 end of the first cycle were only /5 the number remaining in the control. The percent yield as respect Pu was satisfactory, being 93.5%.

EXAMPLE II In order further to study the effect of the additions a single addition was made. One hundred cc. of starting solution was oxidized, a by-product precipitate made, and the filtrate reduced in the same manner as the control. To the reduced solution was added 2 g./l. tartaric acid and a final bismuth phosphate product precipitate made.

Tartaric acid added at this point did not interfere with the MP0; carrying. The percent gamma carried by the final product precipitation was 26.4%, namely more than twice that in the preceding example, and the percent Pu yield, 95.8%. By a comparison of this example with the preceding example, it is apparent that the presence of Hg+ in the preceding example substantially increased the decontamination in the by-product step and also aided in the final productionprecipitation step.

EXAMPLE III Inasmuch as Hg and Ce are regarded as the preferred combination, a series of nine runs were made to determine the optimum weight of Hg and Ce to be used as addition agents. The Hg was used in the by-product precipitation step and added as Hg(NO just prior to precipitation, and the Ce was added just prior to the final product precipitation.

The runs were made on a 230 g. UNH/l. solution containing rather high activity. With the exception of the addition agents, existing procedures were followed for extraction plus one BiPO decontamination cycle.

The Hg and Ce were added in the following weight combinations: (1) 7 /2 g. Hg/l.-l g. Ce/L, (2) 7 /2 g. Hg-5 g. Ce, (3) 7 /2 g. Hg10 g. Ce, (4) 2 /2 g. Hg 1 g. Ce, (5) 2 /2 g. I-Ig5 g. Ce, (6) 2 /2 g. Hg10 g. Ce, (7) 1 g. Hg0.0 g. Ce, (8) 1 g. Hg-l g. Ce, (9) l g. Hg5 g. Ce.

Analysis by counting and a summary of the results were made, and this indicated that increases in decontamination were obtained in all cases. The agents did not atfect Pu carrying. It appeared that 1 g./l. each of Hg and Ce is a satisfactory combination, although both larger and smaller amounts were operative and gave improved decontamination. In general, it is preferred to use the minimum amounts consistent with obtaining the desired decontamination. This conserves materials by not using excesses. Also any possibility of reagent appearing in the recovered Pu is minimized.

EXAMPLE IV While, for various reasons the combination designated as Hg-Ce, is preferred, there are other combinations in accordance with the present invention which may be used. Several studies .were made to develop combinations that would function in a manner somewhat comparable to the Hg-Ce combination in aiding decontamination. Four runs were made in which 1 g. Hg+ per liter was added just prior to the by-product precipitation, and 3.2 g. diglycolic acid, 3.2 g. tartaric acid, 3.0 g. Get and 3.0 g. Mn+ per liter, respectively, added justprior to the final product precipitation.

The source of Pu which comprised the 230 g. UNH/l. type solution containing high fission activity was used as the starting solution. With the exception of the addition agents, standard procedure was followed for extraction plus one BiPO decontamination cycle.

, All the decontamination factors ran low presumably due to the fact that a sedimation had occurred in the starting solution and was incorporated into the runs. However, the data indicated that the Hg-tartaric acid combination and the Hg-Mn combinations compared favorably with" the Hg-Ce combination.

Some of the analytical data and results are tabulated in the following table: a

Data has been presented in the preceding description establishing that decontamination improvement may be obtained by the use of various reagents. In summarizing, it may be indicated that preferred operations are as follows: The addition of Hg++ prior to BiPO precipitation increases carrying of fission (gamma) activity which is beneficial in the by-product step, but may have an adverse elfect in the product precipitation step. However, the use of Ge (probably as a holdback carrier) gives some decontamination improvement in the product precipitation step. Hence, by the combination of agents significant decontamination improvement may be obtained such as by the use of Hg in the by-product step and with Ce+ added in the product precipitation step. This procedure may also be used in conjunction with existing processes in which scavengers, such as Ce-Zr phosphates are used. Data given in the following table shows that the addition agents of the present invention had some additive effect when used in conjunction with a bismuth phosphate process in which scavengers were also used.

Table EFFECT OF ADDITION OF Hg AND Ce ONv DECONTAMINATION (GAMMA ACTIVITY) Grams Hg++/l. Grams Ce+ By- Product Overall added before added before product pptu. Factor by-product prod. pptn. decontam decontaru for pptn. factor factor Cycle 2. 1 10. 3 l0. 3 106 2. 5 5 l0. 3 9. 7 100 2. 5 10.3 6. 1 62. 8 7. 5 1 l2. 6 9. 8 123 7. 5 5 l2. 6 10. 7 135 7. 5 10 12. 6 9. 0 113 1 0 8. 5 12. 3 105 1 1 8. 5 11. 4 97 1 5 8. 5 18. 9 161 5 0 l2. 9 5. 3 64. 1 0 5 4. 9 10. 6 52 0 0 3. 3 8. 2 27 While the process has been described in particular with respect to the phosphate type of by-product and product precipitates, illustrated by alternate bismuth phosphate precipitates since this constitutes the preferred embodiment as well as a type of precipitate frequently encountered in plant processes, our invention is not limited thereto. Similar procedure may be applied in the formation of other precipitates which function in a similar manner under alternate by-product and product precipitation steps. That is, in other instances where precipitates do not adequately decontaminate or otherwise present difiiculties, they may be improved by the application of the principles of our invention.

It may be seen from the foregoing that we have provided certain supplemental additions to by-product and product precipitation steps which produce substantial improvement in the overall decontamination and also permit a better separation of plutonium. The preferred procedure comprising adding a source of mercury in the byproduct step and a source of cerium in the product precipitation. However, as also indicated, other combinations may be employed, such as adding mercury in the by-product step and tartaric acid or comparable organic reagents in the product precipitation.

While mercury has been described as added as the nitrate since nitric acid solutions are being treated and the use of the nitrate eliminates adding other ions to the solution, our invention does not preclude the use of other compatible sources of material such as mercuric phosphate, or the like. Similar comments apply to the particular source of cerium referred to herein as cerium nitrate. Other cerium salts such as cerium tartarate or citrate and the like may be employed in place of, or in addition to cerium nitrate.

The term decontamination factor used herein is a comparative value or ratio derived by dividing the count value, or average for a series of runs, before any particular step by that count value obtaining after the step. An overall factor is the value obtained by multiplying together several individual factors. As apparent from the preceding examples, a cycle comprises a BiPO by-product precipitation coupled with a BiPO product precipitation.

As described above the by-product, product precipitation and other similar steps used such as a BiPO, product precipitation are, per se, the invention of others, the present invention constituting an improvement of existing practices. Hence, the details respecting such steps are not a limitation on the present invention. In the examples described herein the particular details respecting forming the carrier precipitates were the same for both the processing in accordance with the present invention and for the control.

For example, to obtain a MP0,, carrier precipitate while the normality of the environment where the precipitate is formed has been indicated at about 1 N, it may vary substantially as from below 1 N and up to 2.5 N. For example, the choice of a particular concentration of such acid in carrying out the present invention may be influenced by corrosion effect on the particular piece of metal equipment in which the step is carried out. The concentration of bismuth may vary as from .5 gram per liter to 3 or 4 grams per liter. Losses may be slightly higher with the lower amount. Likewise, the concentration of phosphoric acid may vary as from .3 M to l M, around .6 M being a generally satisfactory value. The digestion period and temperature for the MP0,, precipitate, namely the period allowed for the precipitate to gather or otherwise form prior to centrifuging or filtering, may vary from 50 C.l00 C. for one-half to several hours. Likewise, the technique such as rates and manner of additions are not fixed procedures. In general, as already discussed, the use of the smallest amounts consistent with good operation is preferred in order to avoid waste through excesses or for similar reasons. However, the use of larger amounts is not precluded.

The process may be applied to solutions containing Pu from tracer amounts to several hundred grams thereof. Also large volumes of liquids may be treated. The activities of the solutions may be many times higher than the activities usually encountered in the prior art. However, our invention is not limited in these respects as the concentrations suggested are merely guides.

It is to be understood that all matter 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.

We claim:

1. In a process for the recovery of plutonium from aqueous acidic solutions containing plutonium ions and ions of contaminating fission products, which includes the steps of precipitating bismuth phosphate in a solution containing plutonium ions in an oxidation state no greater than four and separating the resulting precipitate and its associated plutonium from the supernatant solution, and the steps of precipitating bismuth phosphate in a solution containing hexavalent plutonium ions together with ions of contaminating fission products and separating the precipitate and its associated fission products from the supernatant solution, the improvements which comprise providing a source of cerous ions in the said solution of plutonium ions in an oxidation state no greater than four prior to precipitating thebismuth phosphate, and provid ing a source of mercuric ions in the said solution of hexavalent plutonium ions prior to precipitating the bismuth phosphate.

2. In a process for the recovery of plutonium from an aqueous acidic solution containing plutonium ions and ions of contaminating fission products, the steps which comprise maintaining the plutonium in said solution in an oxidation state no greater than four, providing a source of cerous ions in said solution, forming a bismuth phosphate precipitate in said solution, separating the bismuth phosphate precipitate and its associated plutonium from the supernatant solution, dissolving the separated precipitate and its associated plutonium in an aqueous acidic solution, maintaining the dissolved plutonium in said solution in the hexavalent state, providing a source of mercuric ions in said solution, forming a bismuth phosphate precipitate in the resulting solution, and separating the precipitate and its associated fission products from the plutoniumcontaining supernatant solution.

3. In a process for the recovery of plutonium from an aqueous acidic solution containing plutonium ions and ions of contaminating fission products, the steps which comprise maintaining the plutonium in said solution in the hexavalent state, providing a source of mercuric ions in said solution, forming a precipitate oi bismuth phosphate in the resulting solution, separating the bismuth phosphate precipitate and its associated fission products from the supernatant solution, reducing the plutonium in the supernatant solution to an oxidation state no greater than four, providing a source of cerous ions in said solution, forming a precipitate of bismuth phosphate in the resulting solution, and separating the bismuth phosphate pre- 10 cipitate and its associated plutonium from the supernatant liquid.

4. In a process for decontaminating an aqueous acidic solution containing plutonium ions and ions of contaminating fission products, the steps which comprise maintaining the plutonium in said solution in the hexavalent state, providing a soin'ce of mercuric ions in said solution, forming a bismuth phosphate precipitate in the resulting solution, and separating the bismuth phosphate precipitate and'its associated fission products from the plutonium-containing supernatant solution.

5. In a process for decontaminating an aqueous nitric acid solution containing plutonium ions and ions of contaminating fission products, the steps which comprise maintaining the plutonium in said solution in the hexavalent state, providing in said solution 1-5 grams of mercury per liter in the form of mercuric ions, forming a bismuth phosphate precipitate in the resulting solution, and separating the bismuth phosphate precipitate and its associated fission products from the plutonium-containing supernatant solution.

References Cited in the file of this patent UNITED STATES PATENTS 2,776,185 Werner et a1. Jan. 1, 1957 2,785,951 Thompson et a1. Mar. 19, 1957 2,799,553 Thompson et a1. July 16, 1957 OTHER REFERENCES Seaborg: The Chemical and Radioactive Properties of the Heavy Elements, Chemical and Engineering News, volume 23, pages 2190-2193 (December 10, 1945).

Harvey: The Actinide Elements and the Chemistry of Plutonium, Nucleonics, vol. 2, No. 4, pages 30-40, particularly page 34 (1948).

Claims (1)

1. IN A PROCESS FOR THE RECOVERY OF PLUTONIUM FROM AQUEOUS ACIDIC SOLUTIONS CONTAINING PLUTONIUM IONS AND IONS OF CONTAMINATING FISSION PRODUCTS, WHICH INCLUDES THE STEPS OF PRECIPITATING BISMUTH PHOSPHATE IN A SOLUTION, CONTAINING PLUTONIUM IONS IN AN OXIDATION STATE NO GREATER THAN FOUR AND SEPARATING THE RESULTING PRECIPITATE AND ITS ASSOCIATED PLUTONIUM FROM THE SUPERNATANT SOLUTION, AND THE STEPS OF PRECIPITATING BISMUTH PHOSPHATE IN A SOLUTION CONTAINING HEXAVALENT PLUTONIUM IONS TOGETHER WITH IONS OF CONTAMINATING FISSION PRODUCTS AND SEPARATING THE PRECIPITATE AND ITS ASSOCIATED FISSION PRODUCTS FROM THE SUPERNATANT SOLUTION, THE IMPROVEMENTS WHICH COMPRISE PROVIDING A SOURCE OF CEROUS IONS IN THE SAID SOLUTION OF PLUTONIUM IONS IN AN OXIDATION STATE NO GREATER THAN FOUR PRIOR TO PRECIPITATING THE BISMUTH PHOSPHATE, AND PROVIDING A SOURCE OF MERCURIC IONS IN THE SAID SOLUTION OF HEXAVALENT PLUTONIUM IONS PRIOR TO PRECIPITATING THE BISMUTH PHOSPHATE.