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Method of Improving Recovery of Neptunium in the Purex Process

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Method of Improving Recovery of Neptunium in the Purex Process United States Patent [is] 3,652,233 Swanson [45] Mar. 28,1972 [54] METHOD OF IMPROVING RECOVERY 3,326,811 6/1967 Healy 23/341 OF NEPTUNIUM IN THE PUREX 3,432,276 3/1969 Reas 23/343 PROCESS OTHER PUBLICATIONS [72] Inventor: John L. Swanson, Richland, Wash. Bruce et al., ed., Progress in Nuclear Chemistry Series 111, [73] Assignee: The United States of America as Process Chemistry, Vol. 3, Pergamon Press, N.Y., 1961, p. represented by the United States Atomic 255. Energy Commission Primary Examiner—Carl D. Quarforth [22} Filed: Dec. 29, 1969 Assistant Examiner—F. M. Gittes [21] Appl.No.: 888,831 Attorney—Roland A. Anderson [57] ABSTRACT [52] U.S.CI 23/341,23/340,23/343 [51] Int. CI BOld 11/00 The recovery of neptunium values from the Purex Process for reprocessing irradiated nuclear reactor fuel is improved by ad- [58] Field of Search 23/340,341,343; 260/644 ding a rate-accelerating material to increase the rate of oxida- tion of the neptunium in the nitric acid solution in the first ex- [56] References Cited traction column. The rate-accelerating material is formed by UNITED STATES PATENTS adding 1-nitropropane to a sodium hydroxide solution to form the aci-form of nitropropane, then adding sodium nitrite and 2,838,366 6/1958 Beaufait 23/343 acidifying the solution by the addition of nitric acid. 2,847,276 8/1958 Butler 23/341 3,004,823 10/1961 Peppardetal 23/341 6 Claims, No Drawings 3,652,233 1 2 METHOD OF IMPROVING RECOVERY OF NEPTUNIUM SUMMARY OF THE INVENTION IN THE PUREX PROCESS J . , „ I have invented an improvement in the process of extracting neptunium values from a nitric acid nuclear fuel feed solution CONTRACTUAL ORIGIN OF THE INVENTION 5 containing neptunium values, including neptunium in the +5 , . .-,.,, • , . „ valence state, with tributyl phosphate wherein there is added The invention described herein was made in the course of, to said feed so,ution a materia, to increase the rate of oxida_ or under, a contract w.th the United States Atomic Energy {ion of neptunium having a +5 valence state t0 a neptunium 1 slon' having a +6 valence state, said material being formed by ad- BACKGROUND OF THE INVENTION ding 1-nitropropane to an aqueous solution containing dilute NaOH thereby forming the aci-form of nitropropane, adding This invention relates to an improvement in a process for sodium nitrite to said solution and adding nitric acid to acidify separating neptunium from uranium, plutonium and fission said solution. The rate-accelerating material when added to product values and more particularly relates to an improve- the nitric acid feed solution, since it increases the rate at ment in the Purex Process for the processing of irradiated 15 which oxidation of the neptunium occurs in the solution, thus nuclear reactor fuels and for increasing the recovery of neptu- increases the amount of neptunium in the +6 valence state nium values therefrom. Neptunium recovery from the Purex available for extraction. My invention is particularly applica- Process is discussed in "Neptunium Recovery and Purification ble in the Purex Process resulting in greatly improved recove- at Hanford" by R. E. Isaacson and B. F. Judson, I & EC rjes of neptunium values. My improvement as applied to the 20 Process Design and Development, Vol. 3, No. 4, Oct. 1964, purex Process consists of adding a material, called a rate-ac- page 296. celerating material, to the Purex Process during the first cycle NP237 not only has utility as a research isotope but also as a separation. This material increases the rate at which the nep- source of Pu238. The latter isotope has become important to tunium values are oxidized from the inextractable +5 valence the general field of space exploration as a heat source for 2$ state t0 the extractable +6 valence state, thus oxidizing more power units. 0f the neptunium present in the feed solution during the Neptunium is produced by either of the following reactions: period of residence in the extraction column. This greatly in- (n, y) (n, 7) (ff) creases the amount of neptunium values which may be ex- V^-1-* XJ23fl ! * Np!" tracted from the acidic feed solution and which may then be 30 separated from the co-extracted uranium and plutonium Tja!_J » ua: > Npa' values by subsequent processing. By my process, the rate-ac- <L7M— celerating material is made by adding 1-nitropropane to a • . ., .,„,2 ,5 .,„-„. v, „,, .„ dilute solution of NaOH to form the aci-form of 1- Since either U ® or U*» is present m most reactors Np^wil nitro r ane> then adding sodium nitrite and acidifyillg the be produced, along with various fission products during fuel 35 so,ution b ^ addition of nitric add The rate.acceleratin burnup. Neptunium exists in the +4, +5 and +6 valence states material is ^ added tQ the first ,e extraction column but is extractable from aqueous solutions with alkyl where it increases the oxidation rate of the neptunium values phosphates only when in either the+4 or+6 state. .. .. .. ., , , . r „ . present in the nitric acid feed solution, thus permitting more The Purex Process is a solvent extraction process used in . ^ , .. , . , . , c , .„ complete extraction of these values and improving the reprocessing irradiated fuel elements fabricated of normal or 40 recQye the f re slightl6 y1 enriche., d uranium. The organi® c solvent employe, d is 30 «.. i.s thereforc e an objec, . t of- thi... s .inventio n t. o .improv e the percent tributyl phosphate (TBP) in normal paraffin _ . _ , „ . .,; , . f j . t-, , • . ... -j recovery orf neptunium values from the nitric acid feed solu- hydrocarbon (NPH). The salting agent is ni ne acid. ^ jn ^ Pur/X Process In the Purex Process, the irradiated fuel elements are dis- solved in nitric acid to form a feed solution. The feed solution 45 DESCRIPTION OF THE PREFERRED EMBODIMENT then enters the first cycle of the solvent extraction system where a gross separation is effected between the fission These and other objects may be achieved by making the products and the uranium/neptunium/plutonium mixture. acid feed solution from 0.005 to 0.02 M in added This is accomplished by counterflow with the TBP organic sol- nitropropane which is made by adding up to 2 M of 1- vent. The fission products remain in the aqueous phase, the 50 nitropropane to an aqueous solution containing up to 2 M uranium/plutonium and neptunium values transferring to the NaOH to foml the aci-form of 1-nitropropane, adding sodium organic phase. In a second cycle of treatment, the uranium nitrite to the solution in a ratio of 0.3 to 0.6 nitrite to 1- and plutonium are separated. The neptunium values separate nitropropane and acidifying the solution by adding sufficient with the uranium from the plutonium and the plutonium is nitric acid so that the final solution contains from about 3 M to further processed to remove fission products. The neptunium 8 M of nitric acid. Because the composition of the resulting and uranium values are separated by contacting the acid product, which is the rate-accelerating material (RAM), is wherein they are contained with an organic extractant which unknown, the amount of material to be added to the acid feed removes the uranium values, leaving the neptunium values in solution is determined by reference to the concentration of 1- the acid solution for further concentration and recovery. 60 nitropropane which would have been present had it not been In order to extract the neptunium values along with the plu- consumed in the formation of the RAM. This is referred to as tonium values from the acid solution in the first cycle, the nep- added nitropropane. Thus sufficient quantity of the solution is tunium values must be oxidized to the +6 valence state. This fed into the first column so that the acid feed solution contains oxidation is presently accomplished in the process by adding about 0.01 M of the added nitropropane. The presence of the HN02 to the nitric acid feed solution which acts to catalyze 55 added nitropropane in the column accelerates the oxidation the neptunium oxidation. However, the Purex Process is a rate of the neptunium present therein, thus causing a high per- continuous process and residence times in the separation centage of the total neptunium present to oxidize to the ex- columns are not sufficient to permit complete oxidation of the tractable +6 valence, permitting extraction of more of the neptunium values present in the column to the extractable +6 neptunium from the acid feed solution into the organic extrac- valence state. Thus a significant percentage of the neptunium 70 tant. The neptunium values are then separated from the urani- values present in the nitric acid feed solution are presently not um and plutonium values which are also extracted and further extracted. The unextracted neptunium values are lost in the purified. aqueous waste solution to the storage tanks, although some The aci-form of 1-nitropropane is prepared by the addition may be recovered later by additional processing of the waste of 1-nitropropane to an aqueous solution of NaOH. Solutions solutions. 75 having concentrations of 1-nitropropane as high as 2 M may 3,652,233 3 .4 be prepared. The amount of NaOH used is the same as, or just k![KAM]/[KNO!] kSilMj u» (mta.) slightly in excess of, that required to convert the nitropropane ; to the aci-form. 6.o6i:v;::::::::~'~~ aixicx t The sodium nitrite may be added to the sodium hydroxide 0.001 1.0X101 10 0.8 solution either before or after the addition of the 1- 5 oloill 11111111111 lioxw loo o?i» nitropropane.
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