Radiochemical Determination of Uranium-237 F. L. MOORE and S. A. REYNOLDS Analytical Chemistry Division, Oak Ridge National Laboratory, Oak Ridge, Tenn.
b A radiochemical method for the determination of uranium-237 is based on complexing the uranyl ion in alkaline solution with hydroxylamine hydrochloride, followed by scavenging with zirconium hydroxide and extraction of the uranium from hydrochloric acid solution with tri(iso-octy1)amine-xylene. The technique has been applied successfully to the determination of uranium-237 in homogeneous reactor fuel solutions.
S
of the formation and chemical behavior of uranium-237 in homogeneous reactor uranyl sulfate fuel solutions necessitated a method for the determination of this isotope. Two pertinent references (4, 10) were found; however, a faster method was needed for routine use with highly radioactive fission product solutions, which are permitted to decay a few days to 1 or 2 weeks after removal from the reactor. Vranium-237 (h/*, 6.75d), a beta, gamma emitter, is formed according t o the following reactions: TUDIES
U-238 (n, 2n) + T;-237
Table 1.
Components of Typical Reactor Solution
Component
Mg. per AIL
Uranium Copper Nickel Sulfate
Concn.,
Inactive
10 1
0.4 7
Radioactive D.P.M. Uranium-237 1x 1x Phosphorus-32 Sulfur-35 4 x 5x Copper44 Strontium-89 3x Y ttrium-9 1 3x Zirconium-95 4x Niobium-95 1x Molybdenum-99 3x 1x Ruthenium-103 6 X Tellurium-132 Iodine-1 31 3 x Iodine-132 6 X Cesium-136 2 x Cesium-137 2x Barium-140 1x Lanthanum-140 1x Cerium-141 4 x Cerium-144 4x Praseodymium-143 4 x Seodymium-147 4 x Xeptunium-239 2x
per hll. 108 107
106 108 109 109 107
107 109 107
lo8 108
lo8
107 107 10'0
10'0 109 108 109
109 109
+ U-236 ( n , ~+ ) U-237 U-235 (n,*,)
The presence of neptunium-239, plutonium-239, and fission products precluded a direct measurement of the uranium-237 by gamma scintillation spectrometry. A typical homogeneous reactor uranyl sulfate solution contains the components listed in Table I and the p H is about 1.3. The method developed involves complexing the uranyl ion in alkaline solution with hydroxylamine hydrochloride followed by scavenging m-ith zirconium hydroxide and extraction of the uranium from hydrochloric acid solution with tri(iso-octy1)amine-xylene (Union Carbide Chemicals Co., New York 17,
N.Y.). PREPARATION A N D STANDARDIZATION OF URANIUM CARRIER
Keigh out approximately 50 grams of uranyl nitrate hexahydrate. Dissolve and make to 1 liter with 2M nitric acid. Standardize (9) the carrier by pipetting 5-ml. aliquots into 50-ml. glass centrifuge cones and precipitating ammonium diuranate by adding concentrated ammonium hydroxide. Filter quantitatively through No. 42 Khatman filter paper and ignite in porcelain crucibles 1080 *
ANALYTICAL CHEMISTRY
at 800' C. for 30 minutes. Weigh as U308.
PROCEDURE
I n a 40-ml. tapered centrifuge tube add 1 ml. of uranium carrier and 0.2 ml. of zirconium carrier (approximately 10 mg. per ml. of zirconium) to a suitable aliquot of the sample solution. Dilute to approximately 10 ml., mix well, and precipitate ammonium diuranate by the addition of concentrated ammonium hydroxide. Centrifuge for 2 minutes and discard the supernatant solution. Wash the precipitate once with 15 ml. of ammonium hydroxide (1 t o 1). Dissolve the precipitate in 1 to 2 ml. of concentrated hydrochloric acid solution, dilute to about 10 ml., add 1 ml. of hydroxylamine hydrochloride (5M), and mix well. Precipitate zirconium hydroxide by the addition of concentrated ammonium hydroxide. Centrifuge for 2 minutes, add 0.2 ml. of zirconium carrier, and stir the supernatant solution, being careful not to disturb the precipitate. Centrifuge for 2 minutes. Add 0.2 ml. of zirconium carrier and repeat. Transfer the supernatant solution to another 40-ml. centrifuge tube, add several drops of phenolphthalein, and adjust the p H to approximately 8 by add-
ing concentrated hydrochloric acid solution dropwise. Add a n equal volume of concentrated hydrochloric acid solution and extract for approximately 0.5 minute with a one-half volume portion of 5% (w./v.) tri (iso-octyl) amine-xylene, Discard the aqueous phase. Wash the organic phase by mixing for 0.5 minute with a n equal volume portion of 5iM hydrochloric acid solution. Repeat the wash step. Strip the uranium from the organic phase by mixing thoroughly with an equal volume portion of O.1M hydrochloric acid solution for 0.5 minute. Discard the organic phase. Add 0.2 ml. of zirconium carrier, mix well, and repeat the above procedure, beginning with the precipitation of ammonium diuranate. Finally, precipitate ammonium diuranate by the addition of concentrated ammonium hydroxide. Centrifuge for 2 minutes. Decant and discard the supernatant solution. Filter on S o . 42 Whatman filter paper and ignite at 800' C. for 30 minutes. Keigh the uranium oxide on a tared aluminum foil (0.0009 inch), fold, and place in a 10 X 75 mm. culture tube. Insert a suitable cork and count the uranium-237 gamma radioactivity in a well-type scintillation counter. Count the same day of the last chemical separation. Apply a blank correction if very low uranium-237 levels are being determined. Determine this correction by taking the same aliquot of uranium carrier through the exact procedure described above. The blank correction is due primarily t o the gamma radioactivity associated with the uranium-235 in the uranium carrier. EXPERIMENTAL A N D DISCUSSION
I n early experiments attempts were made to use a combination of lanthanum fluoride precipitation and ether extraction steps to effect decontamination of the uranium-237. Fluoride chemistry was found unnecessary, after experiments established that excellent separation of the uranium-237 from many elements could be effected by zirconium hydroxide scavenging in the presence of hydroxylamine hydiochloride. The use of hydroxylamine for complexing uranium has been described by Becker and Jannasch (2) and Hecht and Donau (6) and applied by other workers (1, 3, I O ) . A preliminary precipitation of ammonium diuranate removes most of the copper, alkali elements, alkaline earth elements, and anions. It was inconvenient to use the large volumes of highly con-
centrated aluminum nitrate salting agent t o get satisfactory yields in the ether extractions. Therefore, the ether extraction was replaced with the recently developed tri(iso-octy1)amine method (8) which does not require a solid salting agent. While the uranyl-hydroxylamine complex appears strong in alkaline solution, it is rendered ineffective upon the addition of concentrated hydrochloric acid solution prior to the extraction. One cycle of precipitation and extraction steps effected a good decontamination of the uranium-237 but mas erratic, in that small amounts of impurities were detected occasionally. Therefore, the procedure recommended involves two cycles; it yields a uranium-237 product of excellent radiochemical purity when tested on process solutions. Decay studies and gamma scintillation spectrometry were used to check the purity of the uranium oxide product. Typical decontamination data are shonm in Table 11. Gamma spectra of separated uranium-237 were studied to evaluate decontamination factors achieved in the analysis of three reactor samples. This was a precautionary measure to ensure that interference was not caused by some fission product occurring in a chemical state different from that in the tracer solutions described earlier (8).
The scintillation spectrometer employed consisted of a 3 X 3 inch sodium iodide crystal with associated electronic equipment including a 20-channel analyzer. From the spectra of the separated uranium-237, similar to that reported previously (5), the maximum disintegration rates of several fission products were estimated. These disintegration rates were compared with the radiochemical analyses of the samples and the minimum decontamination factors (Table 11) were calculated. Yields averaged 60 to 70 yo. The time required for analysis is approximately 2 hours. The technique previously described (7) of performing extractions in test tubes appeared to be as satisfactory as the conventional use of separatory funnels. The new method has the advantages of speed and elimination of the use of fluoride chemistry as compared with the present method (IO). It should find application in the separation and determination of other uranium isotopes. ACKNOWLEDGMENT
The authors acknowledge the capable assistance of D. K. Smith and P. S. Gouge in testing the procedure. LITERATURE CITED
(1) Bane, R. W., U.
S. Atomic Energy
Table II. Minimum Decontamination Factors Achieved on Actual Samples
Nuclide Mo-99
._ ._ .
Zr-Nb-95 Ru-103 1-131 CS-137 Ba-140 La-140 Ce-141
Decontamination Factors One cycle Two cycles 20 3 . 5 x 102 1 X 102 6 x 102 1 x 102 1 . 5 x 102 2 x 103 1 x 103 1 . 5 ’ X lo4 1 . 5 X lo4 1 x 102
x 102 x 104 x 104 1 x 104
3 2 2
Comm. Declassified Rept., CC-3336 (November 1945). ( 2 j Becker, A., Jannasch, P., Physik. 2. 12.I 1 I19lfi’i. ~
\ - - - -
(3)Brinton, P: H., Ellestad, R. B., J. Am. Chem. SOC.45, 395 (1923). (4) . , Bruce. F. R., Baldwin, W. H., U. S. -4tomic ‘ Energy Comm. Declassified R e d . MonT-199 (October 1946). (5) e e i t h , R. L., Zbid., Unclassified Rept., IDO-16408(July 1957 . (6) Hecht, F., Donau, ., “Anorganische Mikrogewichtsanalyse,”Julius Springer, Vienna, 1940. (7) Moore, F. L., ANAL. CHEN. 2 9 , 448 (1957). (8) Ihzd., 30, 908 (1958). (9) Rodden, C. J., “Analytical Chemistry of the Manhattan Project,” pp. 46-7, McGraw-Hill, New York, 1950. (10) Warren, B., U. S. Atomic Energy Comm. Unclassified ReDt., LA-1721. 244 (September 1954). RECEIVED for review October 6, 1958. Accepted January 7 , 1959.
1
Determination of Uranium and Beryllium in Fused Fluoride Systems N. E. ROGERS and W. D. PRATHER Mound laboratory, Miornisburg, Ohio
b A method is described for the quantitative separation and determination of uranium and beryllium in the presence of macro quantities of sodium in a ternary system of fused fluorides. The fluoride ion is volatilized as fluoboric acid during the dissolution of the dried, mixed salts with acids. Sexivalent uranium is separated from beryllium and sodium by the electrolytic deposition of the uranium as a hydrous oxide on a platinum electrode from a hot ammonium acetate solution a p H 4.0. The beryllium remains in the plating solution with the sodium, and is determined as the hydroxide. Both the uranium deposit and the beryllium hydroxide precipitate are ignited and weighed as their oxides. Sodium is determined by flame photometer or calculated as the difference between the combined weights of the uranium and beryllium fluorides and the original sample weight.
A
determination of the components in the fused salt ternary system, UF4-BeF2-NaF, was required in the determination of the phase relationships of the system. This report discusses a satisfactory analytical procedure which was developed for the quantitative determination of uranium and beryllium in the presence of macro quantities of sodium in such a fused salt system. Previously reported methods for the separation and determination of uranium from beryllium have been limited to a comparatively few gravimetric prccedures. Wunder and Wenger (9) reported the chemical separation of uranium from beryllium by doubleprecipitation of uranium with hydrogen peroxide in chloride solutions followed by the determination of beryllium as the hydroxide. Complexing the uranyl ion with ammonium carbonate ( I ) and determining the beryllium as a basic carbonate N BCCURATE
were tried unsuccessfully in this laboratory. In addition t o giving erratic results, the method was tedious and timeconsuming. Of the general separation techniques, electrolytic methods seemed to offer the most promise and were investigated more thoroughly than other methods. The determination of uranium by electrolysis has been known for many years. In 1880, uranium was reported to be quantitatively electroplated on platinum in a hot ammonium acetate solution ( 7 ) . Kollock and Smith ( 6 ) , in 1901, described the electrolytic separation of uranium from the alkaline earth elements, barium, calcium, magnesium, and zinc. I n spite of the excellent results reported in the early part of the century, additional data on the electrodeposition of uranium have been very scanty during the past 50 years ( 2 , 3). Although electrolysis had not been previously employed specifically for the VOL. 31, NO. 6, JUNE 1959
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