MILDRED J. BRADLEY and LESLIE M. FERRIS O a k Ridge National Laboratory, O a k Ridge, Tenn.
Recovering Uranium from Graphite Fuel Elements This
method, using 90% "03, is suitable as the head-end treatment for the Purex process G R m m m - B A s m FUEL ELEMENTS are being developed for a variety of hightemperature, gas-cooled reactors (6). The proposed fuel elements vary in size, shape, and uranium content, and, in some cases, contain thorium in addition to uranium. T o keep pace with advancing technology, this laboratory is engaged in developing methods for processing the spent fuels. A grindleach process (7), based on leaching mechanically ground fuel with boiling 15.8M (70%) " 0 2 , resulted in lower uranium recoveries than desired with fuels contain less than 5y0 (weight) uranium. Furthermore, grinding of highly radioactive fuel would present formidable engineering problems. Chemical disintegration of graphite fu$l specimens with 21.2M (90%) " 0 3 , bromine (8, g ) , IBr, and IC1 as a substitute for mechanical grinding was investigated. Graphite reacts with these and other reagents to form lamellar compounds (.5, 7, 8, 70, 77). The reactions are usually accompanied by marked swelling and sometimes by physical disintegration of the graphite ( 8 ) . A process based on simultaneous disintegration and leaching with 90% " 0 3 is the most promising of those tested. All results were obtained with unirradiated fuel specimens on a laboratory scale; therefore extensive engineering study will be required before feasibility of the process is established.
oxide on exposure to laboratory atmosphere. The iron concentration in all fuel samples was less than 0.2Y0. Procedure. All experiments were conducted in borosilicate glass with IOto 70-gram samples. Solid-liquid separations after treatment with concenwere made by vacuum trated " 0 3 filtration through medium porosity sintered glass. After its use for disintegration or swelling, bromine was removed from the graphite residue by vacuum distillation. The distillation was begun at room temperature, but ultimately the system was heated to about 100' C. to facilitate removal of sorbed bromine. Four to 6 hours were required for the distillation; IC1 and IBr were removed from graphite residues by distillation a t 150 to 200' C. a t atmospheric pressure. Residues produced by bromine or interhalogen disintegration were leached acwith boiling 15.8M (70%) " 0 3 cording to the method developed for the grind-leach process (7). In all cases, residues were sieved to determine particle size distribution. Tentative " 0 3 Process A process involving disintegration and two leaches with 21.2M (9070) "01 for the recovery of uranium from a 100kg. charge of graphitized fuel containing 5% (weight) uranium is shown below. The conditions given were optimum
with 15-gram samples 1.2 cm. thick. Uranium loss to the graphite residue was less than 0.15% when the fuel contained at least 5y0 (weight) uranium and was greater than 1% when the fuel contained less than 1.5% (weight). I n the first 4-hour leach, graphite pieces began to swell almost immediately upon contact with the acid, increasing three to five times in volume and absorbing about 3 ml. of acid per gram of fuel. This spongy mass crumbled within 1 hour to a porous powder of about twice the original volume. After the second leach, the graphite powder was finer than 20 mesh; 60y0 or more was finer than 100 mesh. Thevolumes of acid and water required for leaching and washing were based mainly on sorptive capacity of the graphite, using vacuum filtration for solid-liquid separation. Suitable feed solutions for tributyl phosphate solvent extraction processes for uranium decontamination and recovery (3) may be prepared by boiling off excess HNO, from the leach solutions and concentrating the wash solutions before they are combined. The acid concentration in the leach solutions was greater than azeotropic (15.2M, 68y0,, maximum boiling mixture) ; therefore, these solutions should be boiled down before mixing with the wash solutions so that condensates, about 21M " 0 3 , can be recycled.
2L2M HNO,
I
U, 5 h g .
C. 9 5 k g I E cm thick
-
I
3 5 0 lilers
3 5 0 liters
-
FIRST LEACH* 4 hc. 93'C.
w
~ 25°C.
T
SECOND ~ LEACH* ~ ~ 4hc, 100°C.
wA ",',: ~
~
--c ~
25'C.
Experimental
RECYCLE ACID 21.2M HN03
FIRST LEACH
Materials. Fuel specimens were made by the same method as the graphitized material employed in the grind-leach expsriments (7). Batches were prepared by admixing UOZ (and sometimes T h O J with graphite flour and a liquid organic binder, pressing the desired shape, removing the binder by baking a t 1000' to 1400' C., reimpregnating with binder, and graphitizing a t 2400' to 2800' C . During graphitization, the uranium was converted to UC2; however, since the specimens were rather porous (density = 1.65 to 2.05 grams per cc., depending on uranium content), the carbide rapidly reverted to the
GRAPHITE RESIDUE ~ C, 9s 5 k g u LOSS, 0 ( 4 %
PRODUCT U, 14 g j l i t e r .
SECOND LEACH 2 3 0 liters.
NITRIC ACID BOIL-OFF
PRODUCT U, 0 . 3 d l i t e r 2 90OF TOTAL. 20.0M "03
I
*EACH S T E P FOLLOWED B Y VACUUM FILTRATION U. 1.95 g/liter 33% OF TOTAL.
F E E D ADJUSTMENT A N D SOLVENT EXTRACTION
3.0 M H N O a 8 4 5 liters.
Process for disintegrating and leaching graphitized 5% uranium-graphite fuels with boiling 90% HNOI Conditions shown were optimum for 15-gram samples 1.2 cm. thick
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Disintegration can be achieved at 25" C., and/or the residue can be washed with 90% H N 0 3 instead of water with only slightly lower uranium recoveries.
was obstinately held in the lattice and was often recovered only after complete destruction of the graphite. Uranium recovery from a specimen containing 7.17, thorium was that expected on the basis of the uranium concentration alone; however, the specimen had been prepared from admixed rather than coprecipitated oxides, a fact which may markedly affect the leaching behavior (2). Acid Concentration and Leaching Time. Maximum uranium recoveries were obtained after two leaches with boiling 90% Hn'Os. Recoveries were lower by 0.27, or more when 7070 acid was used in the second leach (Table I ) . The 1.2 X 1.4 X 4.5 cm. specimens used in this study were completely disintegrated in a 4-hour contact with 30y0 "Os. When the contact time was more than 20 hours, a slurry of very fine particles and some colloidal carbon was produced. No disintegration occurred when the acid concentration was less than about 2 0 M . Leaching Temperature. Specimens containing more than 201, uranium were
Variables Affecting M e t a l Recovery Uranium a n d Thorium Concentrations. Under flowsheet conditions, the uranium recovery decreased with decreasing uranium concentration in the fuel; however, even with fuel containing 0.7% uranium the recovery was 9770 (Table I). The graphite residues contained between 70 and 160 p.p.m. of uranium (average 110 p.p.m,) irrespective of the original uranium concentration in the fuel. The tendency for uranium concentration in the graphite residue to approach a limiting value was also observed by Fromm (4) and is probably due to the formation of the graphite residue compounds analogous to those studied by Hennig (5). Rudorff ( 7 0 ) reported that when a graphite compound was decomposed a small proportion of the intercalated substance
Table 1.
Uranium Recoveries Were Higher when 90%
Was Used
"03
Boiling solution used in each leach
Uranium Recovered, % Fuel Composition,
a
Density, Grams/
90% "08 1st leach
% U
cc.
4 2 water washes
0.63 0.70 2.53 5.41 8.03 10.2 13.0
1.65 1.65 1.83 1.88 2.06 1.90 2.06
91.6 89.1 96.8 98.1 98.2 98.4 99.0
Combustion analysis.
Table II.
3rd water m-ash 0.02 0.1 0.2 0.2 0.1 0.1 0.1
707, "Os 2nd leach Graphite 1 wash residue"
9070 " 0 3 2nd leach Graphite 1 wash residuea
+
4.0 8.4 2.5 1.6 1.6 1.4 0.8
+
3.6 7.4 1.9 1.3 1.4 1.2
4.46 2.3 0.5 0.15 0.08 0.05 0.06
4.9b 2.5 1.1 0.35 0.34 0.33 0.20
0.7
6-hour leaches.
Uranium Recoveries Were Only 2 to 3y0 Lower at 25" C. for Fuels Containing More than 2% U Uranium Recovered, % 0.7% U 25'
90% "03 1st leach water washes 9070 "01 2nd leach wash Graphite residue
+3 +1
2.6% U 25' Boiling C.
Boiling
C. 76.8
90.4
94.1
97.0
11.1 12.1
6.2 3.3
5.5 0.5
2.5 0.5
u
25'
5.4% Boiling
C. 96.6 2.9 0.47
98.3 1.6 0.15
10.2% U 25' Boiling C. 96.4 3.3 0.28
98.5 1.4 0.05
~
Table 111. Uranium in Fuel,
Three Water Washes Were Adequate Uranium Recovered, yo
No.
%
90% "01 1st leach (25' C.)
1
0.72 2.65 5.48 5.30 10.29
49.7 57.8 64.1 65.2 70.6
Run
2 3 4
5 25'
280
c.,90%
"08.
INDUSTRIAL AND ENGINEERING CHEMISTRY
1st water wash
2nd water wash
3rd mater wash
23.8 32.0 29.8 22 .4a 23.6
2.7 3.9 2.4 7.4" 2.0
0.6 0.4 0.3
...
0.2
disintegrated and leached in 4 hours almost as efficiently at 25' C. as at the C . ) . The amount boiling point (-93" of uranium recovered in the first leach and three washes was only 2 to 37, lower a t 25" C. and after the second leach only about 0.291, lower (Table 11). Disintegration and leaching a t 25" C. yielded solutions having the light yellow color typical of dilute UOz(N03)2. Disintegration at the boiling point produced wine-red colored solutions. Species causing the coloration have not yet been identified. Washing. Large quantities of H N 0 3 and UO2(SO?)2solution were sorbed by the graphite, in part through the formation of a graphite-HAT03 lamellar compound. As a result, 25 to ?joy0of the uranium was recovered in the washing steps even when vacuum filtration was used for the solid-liquid separation (Table 111). Three water washes \vere adequate to ensure removal of all uranium solubilized during the disintegration. Water is a more efficient washing agent for soluble uranium than HNOa because it destroys the lamellar compound (70). However, washing with 90% HNO3 is potentially advantageous in that acid added for the second leach is more effective since it is not diluted with water sorbed by the graphite and all solutions are of approuimately the same acid Concentration, thereby simplifying the boildown and recycle operations.
Bromine Disintegration and Swelling The extent of disintegration of graphite fuel specimens agitated in liquid bromine at 25" C. varied primarily with the uranium concentration in the fuel. In 3 to 6 hours, the fraction of a 40-gram specimen reduced to -20 mesh powder increased from essentially zero with electrode graphite to about 957, with fuel containing 870 uranium (Table IV). There was no apparent correlation between disintegration rate and sample density. Increasing the reaction time from 4 to 23 hours produced only a slight further decrease in particle size with fuel containing 0.7% uranium (Table IV,Runs 2 and 3). With fuel containing 970 uranium, a significant reduction in particle size occurred when the reaction time was increased from 1 to 3 hours (Table IV, Runs 6 and 7). Leaching of bromine-disintegrated specimens containing 2 to 9% uranium, according to the procedure described ( 7 ) , with refluxing 15.8M H S 0 3 resulted in recoveries of 99.17 to 99.8670 (Table V). -4lthough the fuel containing 0.770 uranium was only swelled by the bromine, about 3670 of the uranium was recovered by leaching. This recovery is about the same as that obtained wirh boiling 90% HXO3 (Table I)
URANIUM RECOVERY and is markedly higher than the 87YO obtained by leaching -200 mesh specimens of the same material ( 7 ) . Experiments were conducted to determine the effect on uranium recovery 8 mesh samples with of swelling - 4 bromine prior to leaching with 15.8M "03. Bromine treatment of ground samples caused swelling but no disintegration, even with fuels containing more than 5% uranium. Swelling of fuels containing 0.6 to 2y0 uranium resulted in higher recoveries (97 to 99y0, Table V) than obtained by either the grind-leach ( 7 ) or 90% "01 processes. Grinding finer than 8 mesh prior to the bromine treatment was of little value, even with specimens containing less than 1.570 uranium. Bromine swelling had little effect on the recovery from specimens containing
+
5Y0
more than uranium. As in the 9070 HNO: and the grind-leach (7) processes, thorough water washing of the graphite residue after each leach was necessary to obtain maximum metal recovery (Table V).
Disintegration with lnterhalogens
Fuel specimens containing 0.7 to 1401, uranium were readily disintegrated to -20 mesh powder by refluxing for 2 hours in either IC1 or IBr. Disintegration in these reagents was much faster than in bromine but slower than in 90% "01. Leaching of the powders produced by disintegration of fuel containing 0.7% uranium with IC1 and IBr resulted in recoveries of 96 and 99y0, respectively. The latter was the highest
Effect of Uranium Concentration and Reaction Time on the Disintegration of Graphite Fuels in Liquid Bromine at 25" C.
Table IV.
Increased reaction time reduced particle size with fuel containing 9% uranium
No.
Sample Size,a Grams
lb
14
0
1.49
8
2
43
0.70
1.65
4
Run
U in Fuel, %
Density, Grams/Cc.
Reaction ~
i
~Particle ~ Size , Distribution
Mesh
Hr.
%
No disintegration observed
-1- 10
+ 20 - 20 - 10
3
37
0.66
1.65
23
4
43
1.88
1.88
6
+- 1100 + 20 20 +-1010 + 20 -
+ 100
5
37
7.92
1.93
4
6
23
9.31
1.93
1
-20 - 100 10 - 10 -20 - 100 10 - 10 -20 - 100 +20 -20 - 100
+ + 20 + 100 + + 20
+ 100
7
9.20
22
1.93
3
Specimens were about 4 X 4 em. and 0.8 to 1.5 em. thick. cm. in diameter and 12 inches long. a
_
_
_
_
_
_
_
_
Table V.
_
_
_
_
~
~ ~
~
+ 100
88.6 2.9 8.6 86.7 4.9 8.4 9.2 36.3 49.7 4.8 3.7 0.4 68.7 27.2 6.0 21.1 62.5 10.4 0 57.5 42.5
Electrode graphite rod, 0.65
~
~~~
Recovery of Uranium from Bromine Treated Fuels
Thorough water washing after each leach was necessary for maximum metal recovery
Run No. 2" 4a 5" 6a 7Q Sb 9b
u, %
1st leach + 2 water washes
Uranium Found, yo 3rd 2nd leach water + l water wash wash
Graphite residue
89.0 1.3 5.8 3.9 98.1 0.73 0.37 0.83 98.8 0.68 0.29 0.23 99.4 0.06 0.34 0.18 99.6 0.02 0.18 0.14 91.8 0.3 5.1 2.8 98.0 0.3 0.9 0.8 lob 99.2 0.26 0.39 0.16 Ilb 99.2 0.34 0.28 0.16 12b 99.3 0.16 0.25 0.26 99.5 0.11 0.21 13b 0.20 Particle size distribution after bromine disintegration of sample is given in Table IV. Samples were ground to -4 4-8 mesh and swelled with bromine prior to leaching. 0.7 1.88 7.92 9.31 9.20 0.67 1.96 5.38 9.38 11.21 13.85
recovery obtained from fuel containing less than 1% uranium by any method. Recoveries from fuels containing 10 to 14y0 uranium which were treated with IC1 were considerably lower (85 to 97y0)than those obtained by the other methods. This behavior is partially attributed to the fact that halogen (primarily iodine) could not be removed from the graphite and apparently inhibited leaching. Discussion Simultaneous disintegration and leaching in 90% HNO, generally resulted in the highest uranium recoveries from graphite reactor fuels containing 0.7 to 14% uranium. Swelling and disintegration of the fuels occurred more rapidly with 90% HNO, than with the other reagents tested. Since disintegration and leaching are carried out simulprocess, one step taneously in the " 0 3 required in the other methods is eliminated--i.e., mechanical grinding or halogen removal. Uranium recoveries were slightly lower than desired but high enough to make the process useful a t least on an interim basis. The major engineering problems expected in the HNOI process are associated with the solid-liquid separation and with equipment corrosion. However, these problems do not appear to be any greater than those manifested in processes involving the halogens or combustion of the fuels. Acknowledgment The authors thank J. M. Blickensderfer for aid in performing the bromine disintegration experiments. Chemical analyses were provided by the groups of G. R. Wilson and W. R. Laing, O R N L analytical chemistry division. literature Cited (1) Bradley, M. J., Ferris, L. M., Nuclear Sa. C3 Eng. 8, 432 (1960). (2) Bradley, M. J., Ferris, L. M., unpublished data (1960). (3) Bruce, F. R., Shank, E. M., Brooksbank, R. E., Parrott, J. R., Sadowski, G. S., Proc. U. N. Intern. Conf. Peaceful Uses At. Energy, Geneva, 1958 17, 49 (1958). (4) Fromm, L. W., U. S. Atomic Energy Comm. Rept. ORNL-238, 1949. (5) Hennig, G. R., Progr. zn Inorg. Chem. 1, 125-205 (1959). (6) J . Franklin Inst. Monograph No. 7, 1960. (7) Moissan, H., Comfit. rend. 120, 17 (1895). (8) Riley, H. L., Fuel 24, 8 (1945). (9) Rosen, F. D., U. S. Atomic Energy Comm. Rept. NAA-SR-213, 1952. (10) Rudorff, W., Advances i n Inorg. Chem. Radiochem. 1, 223-66 (1959). (11) Thiele, H., Z . anorg. u. allgem. Chem. 206, 407 (1932); 207, 340 (1932).
RECEIVED for review June 3, 1960 ACCEPTED December 28, 1960 VOL. 53, NO. 4
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