Recovery of uranium-234 at high isotopic abundance by chemical separation methods

J. inorg,nucl.Chem.,1966.Vol.28, pp. 2193to 2199. PergamonPressLtd. Printedin NorthernIreland

RECOVERY ABUNDANCE

OF URANIUM-234 AT HIGH ISOTOPIC BY CHEMICAL SEPARATION METHODS

P. E. FIGGINS a n d R. J. BELARDINELLI* Monsanto Research Corporation Mound Laboratory1" Miamisburg, Ohio

(Received 31 January 1966) Abstract--Research quantities of uranium-234of high isotopic purity can be obtained from plutonium238 aged 1 yr or more. A 3-year-old plutonium-238 sample containing 2,%o uranium-234 was processed. The uranium was separated from the bulk of the plutonium by anion exchange in 7-2 N nitric acid. Hexone extraction from a high nitrate salt-low nitric acid medium quantitativelyrecovered the uranium from plutonium and most inert impurities. Anion exchange in both nitrate and chloride media was used for final purification, Nearly 2 g (90 per cent yield) of uranium-234 was recovered, contaminated by less than one part plutonium per million. Emission spectroscopy showed only background levels of other impurities. The isotopic purity was determined to be 99.6 per cent. INTRODUCTION

URANIUM-234 is found in natural uranium at trace levels (57 ppm) and decays by alpha emission with a half-life of 248,000 yr. Research quantities of this isotope were desired to permit study of its basic nuclear properties, and for mass spectrographic standards. Previous efforts to obtain uranium-234 consisted of isotopic separation from natural uranium. Starting with a 4 kg (1 ~o uranium-234) byproduct from the uranium-235 enrichment processes, a 1.5-g sample enriched to 94 ~o uranium234 was recovered by the electromagnetic separators at Oak Ridge National Laboratory.it) With the availability of multi-gram quantities of plutonium-238, an easier and cheaper route to research quantities of uranium-234 appeared feasible. Plutonium-238 decays by alpha emission to uranium-234 with a half-life of 87.5 yr. (2) Calculations show that 100 g of plutonium-238 would produce 776 mg of uranium-234 the first year, and nearly as much each succeeding year. Thus, recovery of gram quantities of uranium-234 is a matter of chemically separating small amounts of uranium from plutonium-238 aged 1 yr or more. An anion exchange process was selected for the initial separation of the uranium234 from plutonium-238. The anion exchange system has been studied extensively (a-5) and is used in the production of plutonium-238, t6) Plutonium (IV) in 7.2-normal nitric acid is strongly absorbed on anion exchange resin while uranium (VI) and most other elements are absorbed only weakly, or not at all. After the uranium * Now at Ling-Temco-Vought, Warren, Michigan. t Mound Laboratory is operated by Monsanto Research Corporation for the U.S. Atomic Energy Commission under Contract AT-33-1-GEN-53. {1) B. HARMATZand R. S. LIVINGSTON,Enrichment of Uranium-234, U.S. Atomic Energy Commission Report Y-660 (September 1950). (2) F. D. LONAOIERand K. C. JORDAN,Half-Lfe ofPu ~3s. U.S. Atomic Energy Commission Report MLM-1291 (1965). ta) j. L. RYANand E. J. W/-m~VemGrrr,Ind. Engng. Chem. 51, 60 (1959). (4) j. L. RYAN,J. phys. Chem. 64, 1375 (1960). (6) A. M. AIKIN, Chem. Engng. Progr. 53, 82F-5F (1957). (6) G. A. BuR~Y, Ind. & Engng. Chem.--Process Design & Dev. 3, 328 (1964). 2193

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P.E. FioGINs and R. J. BELARDINELLI

and the inert impurities such as iron, cobalt, and nickel are washed off with 7"2 N nitric acid, the plutonium is recovered in a highly purified state by eluting with dilute nitric acid. The anion exchange process has the advantage that the U/Pu separation does not depend upon the plutonium being held as plutonium (III) which would be oxidized by radiolysis products produced by the high alpha activity of plutonium-238 solutions. Hexone (methyl isobutyl ketone) extraction was chosen for the second stage because of its selectivity in extracting uranium. <7'8) Uranium (VI) is easily extracted from aluminum nitrate salted solutions while plutonium can be reduced and held as non-extracted plutonium (III). Uranium is easily stripped from hexone with dilute nitric acid, and a nearly pure uranium solution is obtained. Additional purification was desired to remove trace quantities of plutonium and inert impurities. A reported analytical procedure tg) was selected for the final purification step but was modified to obtain separation of the plutonium. Plutonium and uranium are loaded onto anion resin from an aluminum nitrate salted solution and the impurities, especially iron, are washed off. The aluminum nitrate solution is displaced with hydrochloric acid and the plutonium is washed off with hydrochloric acid containing a reducing agent. Dilute acid elutes the purified uranium. ANION EXCHANGE A plutonium sample containing 91.3 g of plutonium-238 was obtained for processing. The sample was approximately 3 years old and should have contained about 2"17 g of uranium-234. The plutonium was present as a mixture of oxide and metal. Plutonium dioxide was dissolved in hot concentrated nitric acid with small amounts of hydrofluoric acid to accelerate the dissolution. The metal was dissolved in 1:1 hydrochloric acid and converted to the nitrate by repeatedly evaporating the solution to a small volume, then diluting with nitric acid. The plutonium was converted to plutonium (IV) by reducing with hydrazine and oxidizing the plutonium (III) with air. A 500 ml ion exchange column (4.0 cm in diameter) with a 500 rrd reservoir was used in the anion exchange process. The column was loaded with about 550 ml of Dowex 1X4 resin (50-100) mesh and backwashed to remove broken resin beads. The resin was conditioned with 1 1. of 7.2 N nitric acid immediately before each run. Feed solution for a typical ion-exchange run consisted of both fresh plutonium feed solution and solutions recycled from a previous run. The feed solution was passed through the column at flow rates of 15-20 ml/min. The plutonium exceeded the capacity of the resin, as indicated by the green colour appearing in the effluent. This overloading was intended to improve the uranium recovery, even at the expense of greater plutonium contamination of the uranium fraction, and to leave the plutonium fraction as devoid of uranium as possible. Two to six litres of 7-2 N nitric acid wash solution was passed through the column to wash off the uranium. The solution volumes were measured in a graduate as collected and the solutions were stored in 1-1. plastic bottles until plutonium and uranium analyses were completed. ,7, j. E. G~LErt, The Radiochemistry of Uranium, NAS-NS 3050 (March 1962). ~8~D. R. MAcK~Nzm,Extraction of Plutonium by Hexone from Aqueous Solutions Containing Nitric Acid, Atomic Energy of Canada Limited, Report CRC-487 (1951). ~,3H. M. OCK~NOENand J. K. FOmEte~N,The Analyst 82, 592 (1957).

Recovery of U"4 at high isotopic abundance by chemical separation methods

2195

Since any residual uranium should be eluted easier than the plutonium, the first portion of plutonium eluate (0.35 N nitric acid) was usually collected and held separate; then the remaining plutonium was eluted, and the solutions were evaporated to less than 1 1. before analysis. Plutonium was determined by gross alpha counting and uranium was determined spectrophotometrically by the 1-(2-pyridylazo)-2-naphthol (PAN) method. (1°) A summary of a typical run and analytical results is given in Table 1. TABLE 1.--SUMMARY OF ION EXCHANGE RUN I I I

Feed

Uranium fraction

Intermediate fraction Plutonium fraction Total

Vol. (i.)

Pu (g)*

U (mg)t

1.60 0.80

24"23 1"805 26.04

241 423 664

Fresh feed Recycle

0.92 0.92 0.94

0.013 0.158 0.757

0 113 279

Resin 50% loaded:~ Resin 85% loaded:l: Resin 90% loaded~, some plutonium breakthrough

0.94 0.95 0.49 0.94 0.45 6.54

0.458 0.338 0.161 0.631 0.353 2.869

150 66 17 31 11 667

7.2 N HNO8 wash 7.2 N HNOs wash 7-2 N HNO8 wash Column backwash 7.2 N HNO8 wash

0.45

1.80

25

0'35 N HNOa

3.68§ 4.13

21.26 23.06 25.93

23 48 715

0"35 N HNOs

Comments

* Plutonium determined by alpha counting. t Uranium determined spectrophotometrically. Estimated from colour.of resin. § Solutions evaporated to less than 1 1. before analysis. The first two effluent solutions collected before the plutonium breakthrough were reasonably low in plutonium. After the breakthrough, washing with 7.2 N nitric acid resulted in decreased plutonium levels until excessive gas bubbles (due to radiolysis) in the column made backwashing necessary. The disruption of the resin bed caused a high plutonium level in the column backwash, but the level went down in subsequent washing. The uranium concentration went through a maximum nearly coincident with the plutonium breakthrough, then decreased as the washing progressed. Interestingly, the uranium concentration also showed an increase in the backwashing step. The mass balance of the plutonium indicated less than 0.5 per cent loss, but this agreement was fortuitous considering possible errors in sampling, ~xo)R. J. BALTISBERGER,Analyt. Chem. 36, 2369 (1964).

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P.E. F~c~3ms and R. J. BFJ.ARI)X~.Lta

diluting and counting. The uranium mass balance showed a net gain of 51 mg or about 7 per cent. The total volume of feed and wash effluent solution collected represents approximately 12 column volumes, of which 4"4 column volumes originated as feed solution. The analytical data emphasize the observation that 8-10 column volumes of effluent were required to wash off at least 90 per cent of the uranium from the column, even at high plutonium loadings. Approximately 115 g of plutonium were processed in five ion e x c h a n g e r u n s ; however, part of the plutonium was recycled through one or more runs. A summary of the five runs and plutonium-uranium separation data are given in Table 2. In general, effluent solutions were those collected during the loading and washing steps. Intermediate solutions were those collected during the change from washing to eluting, and eluates and feed are self-explanatory. TABLE 2.--PLUTONIUMURANIUMSEPARATIONEFFICIENCY Vol.

Plutonium

Uranium

(I)

(g)

(%)

Run I Feed

1.54

39.84

Effluent Eluate

2-35 3.0

1.135 33.13 34-265"

Run II Feed

1.90

33.13

100

0.317

Effluent Intermediate Eluate

3"17 0-21 4"0

0.545 5"62 31.93 38.095

1.6 17.0 96-4 115

0.194 0.038 0.153 0.385

Run III Feed

2.40

26.03

100

0.664

Effluent Intermediate Eluate

6"54 0.45 3.7

1.872 2.797 21.26 25.93

Run IV Feed

0.91

36.16

Effluent Intermediate Eluate

5.79 0.36 2"9

5.853 10.04 18"98 34'87

Run V Feed

0.91

16"04

Efltuent Intermediate Eluate

7.09 0.10 3"4

1.20 0.84 12.31 14.35

100 2.8 83.2 86~0

7.2 10.7 81.7 99.6 100 16.2 27.8 52-5 96"4 100 7.5 5.2 76.7 89.4

(g)

(%)

0.394 0.215 0.317 0.532

0.625 0.067 0.023 0.715

40,4 59.6 100

50.4 9.9 39.7 100

87.4 9-4 3.2 100

0.548 0.919 0-048 0.010 0.977

94.1 4.9 1"0 100

1.978 1.937 0 0.067 2.004

96.7 0 3.3 100

Recovery of U ~u at high isotopic abundance by chemical separation methods

2197

It will be noted that in each run the total uranium found after processing was always greater than the initial value. Subsequent analyses proved the higher value to be more accurate, indicating an unexplained negative bias in the initial analyses. The effect was most pronounced in the first and fourth runs where the feed solution originated from hydrochloric acid dissolution of plutonium metal and contained larger quantities of impurities. Therefore, in calculating the distribution of uranium, the larger value was always used. For plutonium, however, some losses usually occurred in the process and the initial plutonium analysis was used in the calculations. In the anion exchange processing, primary emphasis was placed on separating the uranium from the plutonium, and the data show that uranium losses to the plutonium fraction (eluates) ranged from 60 to 1 per cent. The literature {3) reports distribution coefficients of 3500 for plutonium (IV) and 7.8 for uranium (VI) in 7.2 N nitric acid; therefore, the uranium cannot successfully compete with the plutonium for the active sites on the resin. At the high plutonium loadings it was expected that minimum volumes of wash solution would remove any small amounts of uranium remaining on the column. However, in the first two runs total effluents of 4"3 and 5.8 column volumes removed only 40 and 50 per cent, respectively, with an additional 10 per cent of the uranium found in the intermediate fraction of the second run. It was concluded that the poor uranium recoveries were due to insufficient volumes of wash solution. In the last three runs combined feed and wash effluent volumes of ten or more column volumes consistently gave uranium recoveries of 95 per cent or better. The high plutonium concentrations in many of the effluent solutions were generally due to overloading the column, although incomplete valence adjustment to plutonium (IV) was probably also a factor. In the last ion exchange run, the plutonium loading was purposely light to minimize plutonium contamination of the uranium. Analysis of the uranium product of the anion exchange processing (combined effluent solutions) indicated 1"96 g of uranium or 90"3 per cent of the amount of uranium-234 calculated to be present in the original plutonium sample. An additional 0.30 g of uranium (13.9 per cent) was found in the plutonium fractions, giving a total uranium balance of 2.26 g; this was 4 per cent above the calculated amount. This slight excess is probably due to inaccuracies in the analytical results and error in estimating the age of the original plutonium-238. The uranium product contained 1.02 g of plutonium, so plutonium was reduced by a factor of 107 in the anion exchange process. The uranium product solution was also contaminated with gross amounts of inert impurities, principally iron, cobalt and nickel. The plutonium was precipitated as oxalate and converted to oxide for storage while additional uranium234 grows in. HEXONE EXTRACTION

The combined effluents from the anion exchange process containing all of the separated uranium were combined and evaporated to approximately 60 ml, then diluted five-fold with 0"35 N nitric acid. This concentrated the uranium for ease of handling in subsequent steps and removed much of the excess nitric acid. An equal volume of 2 M aluminum nitrate was added as a salting agent and the resulting solution was used as a feed solution for hexone extraction. A batch counter-current procedure with four extractions and a scrub of the organic phase was used. Ferrous sulphamate was added to the first extraction and to the scrub to reduce and hold the

2198

P . E . Ftt~on~s and R. J. BELARDrSELLI

plutonium as non-extractable plutonium (III). Uranium was stripped from the loaded organic phase by contracting three times with 0"35 N nitric acid in a batch counter-current manner. Uranium analyses indicated essentially complete recovery of the 1.96 g of uranium in the hexone step. Samples of the strip solution were mounted for alpha pulse height analysis and the results showed that 22 per cent of the alpha count was due to uranium-234. After the contribution of uranium to the gross alpha count was subtracted, the plutonium content was calculated to be 2.88 mg (0.15 per cent by weight of the uranium). Thus, in the hexone extraction step, the plutonium contamination was reduced by a factor of 350 with a negligible loss of uranium, and the inert impurities content was reduced considerably. FINAL

ANION

EXCHANGE

PURIFICATION

The uranium solution was evaporated to dryness, then taken up in a minimum of 1 N nitric acid. Portions of this solution containing approximately 150 mg of uranium were adjusted to 1.6 M in aluminum nitrate and 0"3 N in nitric acid by adding solid aluminum nitrate nonahydrate and distilled water. A 20 mm diameter ion exchange column, containing approximately 30 ml of Dowex 1X4 resin (50-100 mesh) was preconditioned with 1"6 M aluminum nitrate in 0.3 N nitric acid, and the uranium feed solution was passed through at a flow rate of 2.0 ml/min. Five column volumes (150 mi) of aluminum nitrate/nitric acid wash solution removed most impurities; then two column volumes of 9 N hydrochloric acid displaced the aluminum solution. Six column volumes of 9 N hydrochloric acid containing 0.04 M ammonium iodide reduced the plutonium to plutonium (III) and washed it off the column. Flow rates of approximately 5 ml/min were used with the hydrochloric acid solutions. The uranium was eluted with 0"1 N hydrochloric acid in a few column volumes. The uranium solution was usually evaporated to dryness and taken up in 1 N nitric acid for ease in slide mounting. CONCLUSION

Emission spectrographic analysis of the final uranium-234 products indicated negligible quantities of all inert impurities. Alpha pulse height analysis indicated that the plutonium contamination had been reduced to 1 ppm (by weight) or less. (At 1 ppm, the plutonium alpha count amounts to 0.23 per cent of the uranium-234 alpha count.) The main radiochemical impurity in the final uranium-234 product was TABLE 3.~ISOTOPIC ANALYSIS Mass

Concentration (wt. ~o)

234 235 236 238

99"63 0-10 0.25 0"02

uranium-232 (with a half-life of 74 yr) apparently originating from the decay of trace amounts of plutonium-236 (with a half-life of 2.85 yr) in the plutonium parent material. The uranium-232 contamination amounted to approximately 25 ppm by weight, or approximately 10 per cent of the total alpha count. The results of mass spectrographic analysis, listed in Table 3, show a uranium-234 isotopic purity of 99.6 per cent.

R e c o v e r y o f U ~s4 at h i g h isotopic a b u n d a n c e by chemical s e p a r a t i o n m e t h o d s

2199

Since the plutonium-238 also contains heavier plutonium isotopes which decay to uranium isotopes, it is of interest to calculate the theoretical isotopic composition of the product uranium. The first two columns of Table 4 list a typical plutonium isotopic composition. The last two columns list the weight and isotopic composition of product uranium from 100 g of typical plutonium after three years. Thus the TABLE 4.--URANItrM-234 ISOTOPIC DILUTION DUE TO ISOTOPIC COMPOSITION OF PLUTONIUM Plutonium composition Isotope (%) P u zSs P u ~39 P u 24° P u 241 P u 242 N p ~a7

78"01 16'81 3"49 0"81 0"14 0"74

Isotope

Uranium product composition* (mg)

(%)

U 234 U 235 U ~se U ~37 U 2ss U ~33

1801 1"410 1"085 3"9 × 10 -~ 7"4 × 10 -4 6"6 × 10 -4

99"862 0"078 0"060 2"2 × 10 -s 4"1 × 10 -s 3"7 × 10 -5

* Based o n 3-yr decay o f 100 g o f p l u t o n i u m .

calculations show that the highest isotopic purity which can be expected is 99.86 per cent uranium-234 with about 0"08 per cent uranium-235 and 0"06 per cent uranium-236, if no uranium contamination was present originally. Imperfect uranium-neptunium and uranium-plutonium separations in the production of plutonium-238 can be expected to leave at least trace quantities of uranium in the plutonium, and this uranium contamination would be recovered along with the uranium formed by radioactive decay. As an example, assuming 0.1 per cent original uranium contamination in 100 g of typical plutonium-238 processed after three years, the isotopic composition of the uranium product would be lowered to 94.8 per cent uranium-234. Of course, once the uranium contamination had been removed in one cycle of uranium-234 recovery, a second recovery cycle should produce uranium-234 of nearly theoretical purity. The isotopic composition of the uranium recovered in this work indicated an extremely low uranium contamination in the plutonium parent material.