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IRWIN R. HlGGlNS Chemical Separations Corp., O a k Ridge, Tenn.
Recovery of Uranium from Calcium Nitrate
. . . by
Continuous Ion Exchange
This type of operation makes it practical to use
4 a process not feasible with conventional fixed bed ion exchange because of low solution-to-resin ratios
4 stainless steel equipment, which
is cheaper than chloride-resistant and more rugged than nonmetallic
Mom
THAN HALF of all the uranium produced in the world is recovered by treatment of the ore leach liquor with ion exchange. The ore is leached with HzS04. Uranium(V1) exists as an anion complex in the dilute (0.3M Sod--) leach and exhibits a strong and highly preferential affinity for anion exchange resin (5, 7, 76). The common method for desorbing the uranium from the resin is to contact it with about 1M nitrate or chloride solutions, slightly acidified. I n these dilute solutions, uranium has little or no affinity for the resin. However, in concentrated (5 to 10M) nitrate or chloride solutions a uranyl anion complex exists which will sorb on anion exchange resin ( 7 , 8, 70-73). This concentration effect in nitrate and chloride solutions makes possible a simple recovery-purification process requiring only water as a regenerant. This feature should make processing of nitrate and chloride solutions attractive, since it offers the same chemical economy as ion retardation (3) and ion exclusion ( 2 ) . I n practice, this economy probably can be achieved only if the solutions are already fairly high in nitrate or chloride concentration. I n most other cases, the cost of adding these salts to dilute solutions would be prohibitive unless the highly salted wastes could be recycled conveniently. Quite extensive study has been made of the uranyl chloride system and its use with a continuous ion exchange contactor (8, 70). By comparison, the study of the nitrate system has been superficial, because it was recognized early that distribution coefficients are not as high and exchange rates are slower, the latter probably because of the lower degree of swelling of the resin gel in the nitrate form (73). However, it is sometimes desirable to operate in the nitrate system just to be able to use stainless steel equipment. This is especially true in radiochemical
processing where the materials of construction must have strength and radiation resistance. Also, a higher degree of purification may be expected in the nitrate system than in the chloride system, since fewer ions have an affinity for nitrate-form anion exchange resin. For example, iron(", a common contaminant, is strongly sorbed in the chloride system but essentially not at all in the nitrate system. Ions which lend themselves to concentrated nitrate processing are thorium, plutonium, uranium, neptunium, protactinium, zirconium, niobium, and americium (4, 75). Separations between these elements are possible in many cases by taking advantage of their differing affinities for the resin under varying conditions of nitrate concentration and acidity in the solution phase. For example, thorium and plutonium are strongly sorbed from 6M "03, while uranium is only weakly sorbed from this solution; but uranium is fairly strongly sorbed from 6 N nitrate salt solutions containing no excess acid. The purpose of this study was to determine what yields and production rates might be expected when recovering uranium from concentrated nitrate salts as a function of the significant variables. The practical production rate for recovery of uranium from nitrate salt solutions was found to be slower than from chloride solutions by a factor of about 8, even after giving the nitrate system the advantage of higher temperature of operation. A 99% yield for uranium was indicated for a 10-footlong bed of 20 to 50 mesh anion exchange resin in a continuous countercurrent contactor, at 60' C., with a feed rate of 250 gallons per hour per square foot of 6.25N Ca(N03)2-0.5N H N 0 3 . For these conditions the production rate corresponding to 0.01 gram of uranium per liter in the raffinate is almost 1 kg. of uranium/hour/equare foot. Increasing the operating temperature
from 52' to 64' C. increased the production rate about 25%. Increasing the C a ( N 0 3 ) ~concentration from 6 N to 7.6N increased the production rate about 25%. Using 20 to 50 mesh resin instead of 16 to 20 mesh increased the production rate about 10%. The uranium was stripped from the resin very effectively with a net water flow only slightly greater than one resin bed void volume. The uranium could be allowed to build up in the bed on top of the concentrated nitrate interface, to at least 50 grams of uranium per liter of solution with a resin loading of 20 grams of uranium per liter, without detrimental effect on losses from the operation. I n making a choice between the nitrate and chloride systems, consideration must be given to the competing advantages of the two systems. The nitrate system gives greater purity, but the chloride system offers higher production rate per unit size of equipment. The nitrate system is more compatible with the stainless steel equipment and the
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RESIN PULSE.
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Arrangement of continuous contactor for recovery of uranium from concentrated Ca(N03)2 solutions Intermediate sampling points in each section a r e indicated
VOL. 53, NO. 12
DECEMBER 1961
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4 Figure 1. Increasing the hold-up time gave steeper concentration gradients in the loading section Resin, 2 0 to 50 mesh Dowex 21 K, 1 3 ml. per minute (about 1 inch per minute in 1-inch contactor): resin loadina. auuroximatelv 1 0 grams of’uranium per liter; feed, 6.1“ Ca52’ C. (N03)2-0.5N “ 0 3 , Superficial solution h‘old-up time in 7.33-foot-lonq loading section: 3.5 minutes. X 7.6 minutes. A 12.5 minutes
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this study (p. 999) was made of l-inchdiameter plastic pipe, with a borosilicate glass pipe section in the reservoir to permit observation of resin movement. Resin was moved by main line water pressure. Plastic ball valves were used in the resin loop. Sigmamotor tubing pumps were used to pump solutions and solenoid valves for solution cut-off. The interface between the concentrated nitrate and water strip solutions was detected with a thermometer, since the feed was warm and the strip water cool. Feed was preheated by pumping through a stainless steel coil immersed in hot water. Sample points were spaced about every foot in the exchange sections. The loading and strip sections were 7.33 and 4.25 feet long.
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chemical systems commonly used in nuclear fuel processing plants. The chloride system may have lower reagent costs but requires more expensive corrosion-resistant equipment, though the latter disadvantage disappears if one can use rubber, glass, and plastics instead of all-metal equipment. These materials are not as rugged or radiationresistant as metals.
Flowsheet Description
Kraus and Nelson have studied the distribution coefficients of practically all the metal ions for anion exchange resin in solutions containing varying concentrations of HC1 (77-73). Similar, but less extensive, studies were made for the nitrate system. I n general, distribution coefficients are not as high and equilibration rates are slower ( 4 ) . The uranyl chloride system was adapted to continuous operation by Higgins and others (8, 70) and has been operated on a production basis by the Japan Atomic Fuel Corp. in an 8inch-diameter, rubber-lined contactor.
Experimental Equipment
The Higgins-type continuous ion exchange contactor (6, 9, 74) used for
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In many applications, it is desirable to use stainless steel because of its strength, corrosion resistance, and radiation resistance. Plastic and rubber-lined equipment is normally used for chloride systems, if not more expensive metallic materials of construction. In many atomic energy applications, nitrate solutions are preferred for compatability with existing equipment and processes. In the present study, Ca(NO3)pHNO3 solutions were made from technical grades of hydrated lime and H N 0 3 , using about 0.5M excess acid. This feed was spiked with crude uranyl nitrate. The chloride feed, for comparison purposes, was made up similarly. The uranium was sorbed on the resin as a complex anion from the 6 to 7 N nitrate solutions. The complex was broken by reducing the nitrate concentration-Le., by using water as the eluting agent. Detecting instrunientation or some kind is required to locate the concentration gradient between the concentrated and dilute nitrate solutions for control purposes. Conductivity is normally used, but instruments of this type were not on hand a t the time of the study. A thermometer was found to be satisfactory for this purpose, since the temperature difference between the cool strip water and the warm feed led to a temperature gradient corresponding to the concentration gradient. The anion exchange resin, with a capacity of 1.0 to 1.2 equivalents per liter, could be loaded to over 100 grams of uranium per liter theoretically if the complex is divalent. Past experience with uranyl chloride indicated that
b Figure 3. Increasing the temperature gave a sharper uranium loading profile
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Feed, 6.2N Ca(NO&0.5N “ 0 8 , 6 5 t o 70 ml. per minute (7- to 8-minute hold-up time in 7.33 foot X 1 inch section); resin, 20 to 5 0 mesh Dowex 21 K, 1 3 ml. per minute .E 52’C. 8 64’ C. I
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EQUILIBRATION
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TIME -HOURS
Figure 2. Increasing the nitrate concentration in solution gave higher resin loading, and raising the temperature gave faster equilibration rates. Apparent distribution coefficient was measured as grams of uranium per liter of resin divided b y grams of uranium per liter of solution
.E 0
1 000
5.7N Ca(NO&-0.3N 3.8N Ca(NO3)2-0.2N
“03,
“03,
65‘ C. 65’ C.
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INDUSTRIAL A N D ENGINEERING CHEMISTRY
5.7N Ca(N0&-0.3N 3.8N Ca(NOJ2-0.2N
“03, “03,
15’ C. 15’ C.
FECT FROM FEED POINT
practical loadings could be over I00 grams of uranium per liter. The primary objective of these runs was to get high recovery, so that high loading and high product concentration were not emphasized, though one point of study was to determine if uranium could be accumulated at the water-nitrate interface without interfering with the loading and stripping. Other factors studied were effect of temperature, hold-up time, resin type and size, and salt concentration on recovery and production rate. Results
Effect of Hold-Up Time on Yield. The most striking effect on yield observed, within the 50' to 60' C. range, was the relatively long contact time required. This seriously hindered high production rates and is a most significant factor when considering the practicality of the process. About 20 minutes' contact time was required for a 99% recovery -e+ about 250 gallons/hour/square foot in a 10-foot loading section, at 10% resin loading (Figure 1). To contrast this with another common ion exchange system, uranyl sulfate ore leach pulp may be fed at 1000 gallons per hour per square foot to give a 99.5% yield with about 1.5 minutes of hold-up, at room temperature and at about 90% of resin loading capacity (5). For sulfate, limiting design features are pressure drop and degree of resin loading. Effect of Temperature on Equilibration Rate and Yield. When trying to extract uranium continuously for the first time from nitrate solutions, it soon became obvious that production rate performance was very inferior to that previously experienced with chloride solutions. Other investigators had noticed a strong temperature effect on rate with related systems (75) so a batch test was made to show the qualitative effect of temperature. One hundred milliliters of Dowex 21K 20 to 50 mesh resin, in the nitrate form, was agitated with 4N and 6N C a ( N 0 J 2 solutions, slightly acidified. Uranium was added equivalent to a total resin loading of 25 grams of uranium per liter. One batch was left at 15' C. and the other heated to 60' to 70' C. At the higher temperature, the equilibration rate was several-fold faster and adequately demonstrated that continuous runs had to be made hot (60' C. is the upper recommended level for extended operation with Type I anion exchange resins). Other observations were that the distribution coefficient was definitely greater for 6 N nitrate, but the equilibration rate was faster for 4N, reaching near-equilibrium in 30 minutes a t 60' to 70' C. (Figure 2). O n the 1-inch continuous contactor two similar runs were made, one at
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F E E T F R O M P E E D POINT Figure 5. Smaller particle size resin gave somewhat steeper uranium loading profile than larger size
Figure 4. Higher nitrate concentration in solution gave steeper uranium loading profile on continuous anion exchange contactor
Resin rate, 1 3 ml. per minute; feed, 6 . 3 N "08; solution hold-up time, Ca(NO&-O.SN about 8 minutes 0 2 0 to 50 mesh Dowex 21 K, 64' C. 2 0 to 50 mesh Permutit SK, 57' C. 12 to 20 mesh Dowex 21 K, 6 3 ' C.
Temperature, 5 2 ' C.; resin, 20 to 50 mesh Dowex 2 1 K, 1 3 ml. p e r minute; resin loading, about 10 grams of uranium per liter 0 6.ON Ca(NO&-O.JN "Os, 7-minute hold-up time 7.6N Ca(NO&-0,5N "03, 6-minute hold-up time
Dowex 21K, as might be expected. For the coarser resin, about a 10% increase in column length was indicated to be required to get the same performance (Figure 5).
about 52' C. and the other at 64' C. There was about a 25% increase in effectiveness of the loading section at the higher temperature (Figure 3). Effect of Nitrate Concentration on Yield. The higher the Ca(NO& concentration the higher is the distribution coefficient of uranium for the resin. (A maximum was not found, though in " 0 3 it is about 8 N ) . As shown in Figure 2, the equilibration rate was faster the lower the nitrate concentration. Since the equilibration rate is so significant in this system, a pair of contactor runs were made to see if recovery would be better or worse at higher concentration. Increasing the Ca(N0s) 2 concentration from 6.ON to 7.6N increased loading section recovery by about 25% (Figure 4). Effect of Resin Type and Mesh Size on Yield. An attempt was made to determine if there was significant differences between different particle sizes of the same resin or between similar resins made by different companies. I t was decided that there was no great difference between 20 to 50 mesh Dowex 21K and 20 to 50 mesh Permutit SK. The 16 to 20 mesh Dowex 21K was somewhat inferior to 20 to 50 mesh
from anion exchange resin loaded from concentrated nitrate or chloride solutions is very efficient. At concentrations below 2 N the uranium exhibits little affinity for the resin. From highly loaded resin in the chloride system product solution has been withdrawn at over 200 grams of uranium per liter, using a net volume of water about equal to the resin void volume. I n this nitrate study, the primary concern was to get high recovery from the nitrate salt, with high product concentration secondary. There seemed to be a possibility, however, that the uranium product concentration could be built up by very high facto-s wi.hout interfering with the primary extraction operation. If the resin rate was such that it was loaded from the feed solution well under its maximum capacity, the uranium which accumulated at the water-nitrate interface would merely tend to go back on the resin without overloading the feed section. Frequent samples were taken at a point above the feed but well below the water-nitrate interface, and in every instance the uranium present was either the same or less than in the feed
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Product Elution Efficiency Using Water. The water elution of uranium
VOL. 53, NO. 12
DECEMBER 1961
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Figure 6. Uranium was effectively stripped from resin with only slightly more than one void volume of water. Product uranium concentration was as high as 50 grams per liter with resin loaded to 15 to 20 grams per liter from feed containing 3.2 grams per liter Resin, 2 0 to 50 mesh Dowex 21K, 1 3 ml. per minute; net strip water rate, about 7 ml. per minute or less, controlled by thermometer in strip section
sample, indicating that little or no uranium was being scrubbed down into the loading section. No attempt was made to push product concentration to a maximum, but 50 grams of uranium per liter of product was withdrawn in a run with a feed concentration of 3.2 grams of uranium per liter, The product was stripped with slightly over one resin void volume of water (Figure 6). Because exchange rate was the production limiting factor, the resin was purposely loaded well below its maximum capacity for uranium to keep the exchange potential as high as possible. The straight line shapes of the loading concentration profiles, for resin loadings all the way from 5 to 20 grams of uranium per liter, indicated that the exchange rate was constant. In some ion exchange operations it is desirable to load the resin to near its full capacity to use the chemical stripping agent economically. The loading profile in these circumstances is chgracteristically curved. Because of the effective and low cost method of stripping
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FEET FROM FEED POINT Figure 7. Slope of uranium loading profile was approximately the same for low and high resin loadings Resin rate, 1 3 ml. per minute; feed, 6.ON Ca(NOJ2-0.5N “03, 52’ C.1 hold-up time about 7 minutes 0 Resin loading, 9.6 grams per liter .f. Resin loading, 20 grams per liter
with water, there is no great incentive for pushing for high uranium loading on the resin (Figure 7). Comparison of Yields a n d Production Rates. The difference in the exchange rate of uranyl chloride and uranyl nitrate anion complex is very striking. Even at a lower temperature (30” us. 53” C.), a much higher resin loading (25 us. 5 grams of uranium per liter), and a much lower contact time (3.1 us. 20 minutes), the uranium recovery was more efficient in the chloride system than in the nitrate system (Figure 8). O n this “handicap” basis, production rate for the same yield is about 8 times faster in the chloride system. Thus, it is questionable if the advantages claimed for nitrate (less corrosion and utilization of stainless steel) can overcome the disadvantage of slower production rate.
literature Cited (1) Bunney, L. R., Ballou, N. E., Pascual, J., Foti, S., U. S. Naval Research and Develop. Lab. Rept. TR-228,1958. (2) Dow Chemical Co., Midland, LMich.,
