Iron EDTA chelate catalyzed oxidation of uranium

Uranium ore deposits which contain uranium in the relatively insoluble tetravalent state are readily selectively leached in situ to recover relatively pure uranium compounds, by: (a) passing through the ore deposit a relatively dilute aqueous leach solution of ammonium bicarbonate, ferric ammonium ethylenediaminetetraacetic acid (EDTA), and a source of oxygen, the leach solution converting the tetravalent uranium to hexavalent uranium which readily dissolves in the leach solution; (b) withdrawing the reacted leach solution enriched in dissolved uranium from the ore deposit; and (c) stripping the uranium from the withdrawn leach solution. The stripping of the uranium from the leach solution is preferably accomplished by countercurrent flow of the enriched leach solution to a column of base anion exchange material which preferentially extracts the uranium. Base anion exchange material loaded with uranium is separated from the leach solution and is treated with an aqueous alkaline eluant to extract the uranium and to regenerate the base anion exchange material. The stripped leach solution is adjusted by adding ammonium bicarbonate, peroxide, and ferric ammonium EDTA, and its pH corrected if necessary, and the leach solution is recycled through the ore deposit. The uranium bearing eluant is then acidified and treated with ammonia to precipitate relatively pure ammonium diuranate (ADU).

CROSS-REFERENCE TO RELATED APPLICATION 
This invention is related to and is an improvement upon the invention set 
forth in copending application Ser. No. 513,445 filed Oct. 9, 1974, now 
U.S. Pat. No. 4,155,982, entitled "In Situ Leaching And Recovery Of 
Uranium From Ore Deposits." 
BACKGROUND OF THE INVENTION 
1. Field of the Invention: 
This invention relates to the recovery of uranium from underground ore 
deposits or bodies by in situ leaching and the subsequent processing of 
the enriched leaching solution to recover relatively pure uranium 
compounds therefrom. 
2. Description of the Prior Art: 
Efforts have been made in the past to recover mineral values from 
underground ore deposits by introducing various leaching solutions in 
order to avoid the costs and problems of mining, such as are involved in 
tunneling, blasting and hauling of ore to the surface and then processing 
the ore by various means as by grinding, ball milling and flotation, 
followed by chemical solution or pyrometallurgy to recover the desired 
minerals therefrom. The application of leaching solutions of various types 
to underground ore deposits has been attempted with results that have 
varied widely, and only a few have been particularly successful, as for 
example, in the recovery of sulfur and salt. One of the problems leading 
to a lack of success for leaching out other minerals has been the fact 
that such other mineral whose recovery by in situ leaching is desired is 
that they often comprise only a small proportion of the total volume of 
the soluble minerals and insoluble gangue in the underground ore body. 
Consequently, the leach solutions must penetrate deeply into masses of 
gangue for a small recovery of the desired mineral values. In addition, 
the leaching solutions quite often have reacted with or been contaminated 
by numerous other minerals than the ones particularly desired as well as 
by clays and salts. This arises because the contaminants have also been 
only too well dissolved by the leaching solution. Leach solutions so 
contaminated have necessitated much subsequent refining processing in 
order to separate effectively the desired mineral from the undesired 
materials. A third factor is that excessive amounts of expensive leaching 
materials are necessarily employed because large proportions thereof are 
either dissipated, as by reaction of leach acids with limestone or 
calcite, or else substantial volumes of the expensive leaching solutions 
escape or are trapped and lost in the crevices of the ore deposit and 
never recovered. 
These problems of in situ leaching are particularly critical in the process 
of recovery of uranium which is present in small percentages in most ore 
deposits that are reasonably amenable to leaching in situ. 
U.S. Pat. No. 2,738,253 issued Mar. 13, 1956, discloses an initial 
application of an aqueous solution of sodium chlorate to a uranium bearing 
ore body followed by an acid leaching solution, which latter may or may 
not have additional sodium chlorate present therein, in order to recover 
the uranium values. The inventors in this patent indicate the fact that 
these ore bodies are often associated with ferrous iron along with 
tetravalent uranium. Tetravalent uranium is relatively insoluble in the 
leaching solution. By employing the sodium chlorate, the patent teaches 
that oxidation of the ferrous iron to ferric iron and the tetravalent 
uranium to hexavalent uranium is accomplished so that the acid leaching 
solution will readily dissolve the uranium and render it available. 
Other acid leaching solutions are known, as in U.S. Pat. No. 3,309,141 
issued Mar. 14, 1967, which discloses the combination of sulfuric acid and 
sodium chlorate in a leaching solution for extracting uranium from uranium 
bearing ore. U.S. Pat. No. 3,309,140 issued Mar. 14, 1967 teaches the use 
of a leaching solution comprising from 5 to 25 grams per liter of nitric 
acid and from 0.5 to 2 grams per liter of sodium chlorate. It is taught 
that the sodium chlorate is employed in order to oxidize the tetravalent 
uranium to the more soluble hexavalent uranium ion. Chlorates and nitric 
acid are both relatively expensive and have other drawbacks due to their 
highly corrosive effects on metal valves, piping, etc. 
A number of patents have disclosed the employment of sodium carbonate 
solutions for extracting uranium from underground deposits by a leaching 
operation. U.S. Pat. No. 2,964,380, issued Dec. 13, 1960 discloses the 
general concept of a leachant comprising a 3% sodium carbonate solution in 
water which when applied to crushed uranium ore will leach the uranium 
therefrom. 
U.S. Pat. No. 2,896,930, issued July 28, 1959 states generally that an 
aqueous solution containing "less than 50 grams per liter of dissolved 
carbonates" is suitable for underground leaching of uranium ore. An 
"alkali metal carbonate" is mentioned as suitable for such leaching 
utility. This patent states generally that "It is advantageous to 
incorporate an oxidizing agent such as hydrogen peroxide in the leach 
solution." No specific data or any specific proportions of suitable 
compositions are given in this patent, other than the above quoted upper 
limit for unspecified carbonates. At the bottom of column 3, of this 
patent, it is suggested that the recovery of the uranium whether from the 
leaching solution or from an inorganic solvent into which it has been 
incorporated by solvent extraction, may be effected using an ion exchange 
resin. 
Another patent disclosing the use of carbonates is U.S. Pat. No. 2,818,240 
issued Dec. 31, 1957. This patent discloses that carbonate solutions 
comprising 5 to 14% of sodium carbonate, 2% sodium bicarbonate and 5% of 
sodium chloride form aqueous solutions that would be of a pH of 9.9 to 
9.6, but that the sodium chloride reduces the pH to 9.3. This patent also 
teaches that aqueous solutions of a pH of 9.6 or slightly in excess are 
effective in leaching out more of the various carbonaceous materials in 
the ore deposit. The patent also teaches that the sodium bicarbonate 
depresses the pH, and then it states, "which is undesirable" to secure 
maximum leading of carbonaceous material as is desired. U.S. Pat. No. 
3,708,206, issued Jan. 2, 1973, teaches the pumping of an oxygen bearing 
gas such as air into a uranium ore body in order to oxidize the uranium to 
the hexavalent state, and after many hours or days of exposure to the 
oxidizing gas, a leach solution of sodium carbonate or ammonium carbonate 
is pumped into the oxidized ore body. The patent teaches as desirable 
leaching solutions, those containing from 23 to 26 grams per liter of 
ammonium carbonate. 
U.S. Pat. No. 3,792,903 teaches the recovery of uranium from underground 
ore bodies by introducing leachants comprising sodium carbonate and an 
oxidant which latter may comprise air, oxygen or hydrogen peroxide. No 
specific solution compositions are given except that the patent states 
that the sodium carbonate leaching solution to the oxidizing solution may 
be proportioned from 1:1 to 1:10 by volume. 
U.S. Pat. No. 3,130,960, issued Apr. 28, 1964 teaches the use, as a 
leaching solution, of carbon dioxide gas impregnated water applied to ore 
deposits of uranium and vanadium. It is noted that such leaching solutions 
should comprise at least 20% of the maximum possible carbonation in which 
100% equals 30 volumes of carbon dioxide per volume of water. These 
solutions are obviously acidic. Thirty volumes of carbon dioxide in one 
volume of water provides approximately 59 grams per liter of carbon 
dioxide, while 20% carbonation introduces about 12 grams of carbon dioxide 
per liter. This last patent also teaches that the leach solution, after it 
has passed through the ore body and brought to the surface, is treated 
with lime to precipitate the uranium and vanadium values. 
From the above, it will be apparent that the leaching solutions have 
generally been relatively concentrated and have comprised either acids or 
alkali metal carbonates. U.S. Pat. No. 2,818,240 is the only patent that 
employs a bicarbonate, namely sodium bicarbonate, in a leaching solution. 
None of the references teaches the use of ammonium bicarbonate and none 
suggests employing dilute ammonium bicarbonate solutions, alone, or with a 
peroxide, for leaching uranium values from ore deposits. 
The following articles, comprising papers presented at Geneva, Switzerland 
from September 1 to September 13, 1958 as part of the "Proceedings of the 
Second United Nations International Conference on the Peaceful Uses of 
Atomic Energy," published in Volume 3, "Processing of Raw Materials," are 
of interest with respect to the present invention: 
1. "The Role of Process Development in Western United States Uranium 
Procurement" by J. W. Barnes--pages 183 to 190; 
2. "Some Variations of Uranium-Ore Treatment Procedures" by E. A. Brown et 
al--pages 195 to 200; 
3. "Kinetics of the Dissolution of Uranium Dioxide in Carbonate-Bicarbonate 
Solutions" by W. E. Schortmann and M. A. DeSesa, pages 333 to 344; and 
4. "Extraction of Uranium from Solutions of Sodium Carbonate by Means of 
Anionic Exchange with Dowex Resin" by M. Urgell et al, pages 444 to 464. 
Also note the article by F. A. Foward in the October, 1953 issue of the 
Canadian Min. and Met. Bulletin, entitled "Studies in the Carbonate 
Leaching of Uranium Ores." 
However, none of this last-mentioned art discloses the use of dilute 
ammonium bicarbonate and peroxide leaching solutions for recovering the 
uranium. 
SUMMARY OF THE INVENTION 
The present invention relates to an improved process and leaching solutions 
for extracting and recovering uranium from ore deposits in which it 
resides in its relatively insoluble tetravalent state, by preparing a 
dilute aqueous alkaline leach solution of from about 0.3 to 5 grams per 
liter of ammonium bicarbonate, and preferably a mixture of ammonium 
bicarbonate and ammonium carbonate, a catalyst comprising ferric ammonium 
EDTA in an amount to provide from about 3 to 100 ppm of iron, with the 
EDTA:Fe ratio being at least about 1.25:1, and oxygen provided in an 
amount of from 0.1 to 3 grams per liter of H.sub.2 O.sub.2, or gaseous 
oxygen, the pH of the solution being from about 7.4 to 9.5, a small amount 
of (NH.sub.4).sub.4 EDTA being present to stabilize the ferric ammonium 
EDTA, for example about 1% to 2% of the weight of the latter, and then 
passing the solution through the ore deposit where the leach solution 
first converts the tetravalent uranium it contacts to the hexavalent state 
and then dissolves the hexavalent uranium. The leach solution dissolves 
the uranium preferentially to other contaminating metals and elements with 
which it may be associated in the ore deposit, so that the solution will 
contain a smaller proportion of such elements as molybdenum, vanadium, 
copper, arsenic and selenium to the uranium than their ratios in the ore 
deposit. 
The leach solution enriched in uranium after its passage through the ore 
deposit, is withdrawn from the ground and the uranium is stripped 
therefrom, preferably by passing it countercurrent to a column or bed of 
base anion exchange material such for example as a particulate ion 
exchange resin. The uranium loaded base anion exchange material is 
separated from the leach solution and is treated with an aqueous eluant to 
extract the uranium therefrom, and to regenerate the anion exchange 
material for recycling with more enriched leach solution. The uranium rich 
aqueous eluant is treated first with an acid and the ammonia to 
precipitate ammonium diuranate. It has been found that the contaminating 
metals and elements decrease during the anion exchange material treatment 
of the leach solution as well as during elution and precipitation of the 
ADU, so that a relatively pure ADU product is the final result. 
The stripped leach solution is then treated by adding more ammonium 
bicarbonate and H.sub.2 O.sub.2, as well as the iron chelate catalyst, and 
recirculated to the ore deposit to recover more uranium. 
The ferric ammonium EDTA functions as a catalyst to promote a more rapid 
conversion of the tetravalent uranium, and further has enabled a more 
complete recovery of the uranium present in an ore deposit. 
The leaching solutions contain no components that are harmful or 
deleterious to the underground areas or any aquifer associated therewith. 
All of the components, except the ferric ammonium EDTA, decompose to 
water, carbon dioxide and oxygen. The ferric ammonium EDTA is present in 
very dilute concentrations to begin with, and is a widely used component 
in fertilizers so that it is considered ecologically acceptable. Above 
ground, the process of this invention results in very low quantities of 
impurities so that minimal amounts of waste solutions or solids result 
which require disposal. 
After the ore deposit has been leached out to extent it is economically 
feasible with respect to the uranium present therein, in order to 
insolubilize any residual uranium and other metals and elements that were 
affected by the alkaline leach a reducing solution is introduced into the 
ore deposit, for example an aqueous hydrogen sulfide or ammonium 
thiosulfate solution is passed therethrough, so as to render the uranium 
and other metals or elements as insoluble as they were originally. An 
ecological equilibrium is reestablished with the added advantage that most 
of the uranium is no longer present in that deposit. The process is 
ecologically and environmentally highly advantageous because it produces 
above ground a very low volume of impurities and wastes requiring 
disposal, while underground it causes no significant or material 
contamination of the earth or any aquifer associated with the ore deposit.

DETAILED DESCRIPTION OF THE INVENTION 
The present invention is particularly adapted to recover rapidly and 
efficiently practically all the uranium in an ore deposit wherein the 
uranium is present in the highly insoluble tetravalent state and is 
usually associated with one or more contaminating metals and elements such 
as molybdenum, vanadium, selenium, arsenic, and copper as well as iron. 
These contaminating metals and elements are undesirable in uranium for 
nuclear reactor purposes and it is highly beneficial to reduce their 
presence to small fractions of one percent, so that purification of the 
uranium in subsequent treatments prior to enrichment is easier and more 
economical. The invention is based on the preparation of dilute aqueous 
alkaline leach solutions which are passed through uranium ore deposits and 
solubilize the uranium by converting it to the hexavalent state which 
dissolves readily in the leach solutions in preferential proportions as 
compared to the contaminating elements and metals in the ore deposit. 
Representative of ore deposits that are amenable to the practice of this 
invention are those designated as roll front deposits that are 
characterized by a redox interface therein. 
Briefly geological studies have established the fact that uranium which has 
been brought to the surface of the earth by volcanic action or the like 
will dissolve to some degree in surface waters containing oxidizing agents 
along with carbonic acid. The streams containing the dissolved uranium, 
along with other metal values, may flow into fluvial sand deposits or, in 
some cases porous sandstones, which are quite often overlaid and underlaid 
by mudstone layers of a low permeability. Such sand or sandstone deposits 
may comprise quartz or silica sands, feldspars and often include carbonate 
minerals such as calcite and have varying degrees of porosity and/or 
permeability. The sand or sandstone deposits may include varying amounts 
of carbonaceous matter such as wood or plant residues, and iron sulfide. 
Sometimes oil or hydrocarbon gas deposits underly these sandstone and the 
oil or gas may slowly percolate therethrough. At or in certain areas, 
hydrogen sulfide which may be present further underground may diffuse or 
leak into the sandstone deposits to produce a reducing condition while the 
calcite and feldspar tend to produce a basic condition in the sandstone. 
When the flowing waters containing uranium and other dissolved minerals 
enter such sandstone deposits in which reducing and non-acidic conditions 
are present, the uranium is reduced to the tetravalent state and 
immediately precipitates within the interstices of the sand or sandstone 
formation. Vanadium, molybdenum, and selenium as well as other elements 
are reduced at the same time and also deposit, either concurrently with 
the uranium or nearby. 
Referring to FIG. 1 of the drawings, there is illustrated an exemplary 
cross-section through an ore roll front deposit showing a bullet-shaped 
redox interface, having a vertical tangential redox interface at the head 
of the oxidized zone at the center of the Figure. In the area of the redox 
interface existing between an oxidizing zone through which the waters 
containing dissolved uranium and other minerals enter from the left and a 
reducing zone to the right where the waters contact the reducing materials 
in the sand or sandstone, the uranium precipitates predominantly in the 
reducing zone over a period of time to produce a relatively concentrated 
ore zone in the sand or sandstone interstices. Other minerals in the water 
stream also precipitate either just before or after the uranium deposits. 
The mineral depleted waters then traverse to the right along the 
hydrological gradient through the reduced zone and thereafter disappear 
from the area. In the oxidized zone at the left in FIG. 1, the sandstone 
is brown or red colored indicating the presence of precipitated ferric 
oxide (Fe.sub.2 O.sub.3 or FeO(OH)). A relatively sharp interface exists 
between the oxidized zone and the reducing zone which is evidenced by the 
abrupt change in color in the reduced zone to a light gray or drab color 
characteristic of the reduced minerals. The usual thickness of the 
sandstone containing these minerals is of the order of 10 to 50 feet in a 
vertical direction. The lateral extent of an ore deposit may be hundreds 
or even thousands of feet. The horizontal distances with respect to the 
vertical tangential redox interface are typically indicated in FIG. 1. 
This mineral containing sandstone is usually confined between upper and 
lower mudstones or relatively impervious shales. It is useful to know the 
general order and extent of deposition of some of the more important 
elements with respect to the redox interface and these are shown in FIG. 2 
for a typical roll front deposit found in Texas. It will be noted that the 
uranium precipitates almost precisely beginning at the redox interface, 
with little being present in the oxidized zone, while in the reduced zone 
the uranium concentration drops steadily so that the uranium deposition is 
practically all concentrated within roughly the first 500 feet of entry 
into the reduced zone from the redox interface. In a Texas deposit in the 
Catahoula formation, the maximum concentration of the uranium is about 
2000 ppm at the redox interface and the concentrations drops in a nearly 
straight line to a 10-20 ppm value at the 400-450 foot point away from the 
redox interface. 
As is further evident in FIG. 2, the vanadium is more broadly distributed 
for a considerable distance in the oxidized zone ahead of the redox 
interface, reaching concentrations of about 800 ppm at the redox interface 
and then dropping slowly to a lesser concentration extending for some 
distance forward into the reduced area. Selenium is practically all 
precipitated in the oxidized zone in some 200 feet just immediately before 
the redox interface. Molybdenum, on the other hand, does not appear to be 
precipitated until the water stream bearing it had passed several hundred 
feet from the redox interface into areas of strongly reducing conditions. 
It is also significant that most of the molybdenum is present as 
molybdenum disulfide. Both selenium and molybdenum reach peak 
concentrations of about 200 ppm. Ferrous iron is present in greatly 
varying proportions throughout the area of the reduced zone where the 
sandstones are generally alkaline in nature having a buffered pH value of 
about 8. 
Due to the presence of substantial amounts of calcite and other alkaline 
earth carbonates in roll fronts, the use of acidic leaching solutions 
often is not economically feasible. In many such sandstones the acids will 
react with the various alkaline earth metal carbonates such as magnesium 
and calcium carbonates, before they can start dissolving uranium. 
Consequently, much acid will be lost in ore deposits having substantial 
amounts of dolomite and limestone. 
In accordance with the present invention, it has been discovered that ore 
deposits having characteristics such as those illustrated in FIGS. 1 and 2 
of the drawings, may be rapidly and economically leached to recover a high 
proportion of the uranium in a relatively pure condition and only meagerly 
contaminated with other elements. In particular, it has been discovered 
that relatively dilute leach solutions containing from about 0.3 to 5 
grams per liter, an optimum being from 0.5 to 2 grams per liter, of 
ammonium bicarbonate (NH.sub.4 HCO.sub.3), from about 0.1 to 3 grams per 
liter of H.sub.2 O.sub.2, an optimum being 0.3 to 1.5 grams per liter, or 
oxygen gas or even air, to provide an oxidizing agent, and a catalyst 
comprising ferric ammonium EDTA (FeNH.sub.4 EDTA) wherein the EDTA:Fe 
ratio is at least about 1.25:1, in an amount providing from about 3 to 100 
ppm of iron, an optimum being from about 10 to 20 ppm, the pH of the leach 
solution being from about 7.4 to 9.5, when applied to the ore will rapidly 
and efficiently convert the tetravalent uranium it contacts to the 
hexavalent state which readily dissolves in the leach solution. A small 
amount of the order of one or two percent of the weight of the ferric 
ammonium EDTA, of ammonium EDTA ((NH.sub.4).sub.4 EDTA) is desirably 
present in the leach solution in order to stabilize the iron chelate 
catalyst in the solution. 
There is a preferential leaching of the uranium by the leach solution with 
respect to the contaminating metals and elements in the ore deposit, so 
that much smaller proportions of the molybdenum, vanadium, copper, 
selenium, arsenic are present in the leach solution as compared to their 
proportions in the ore. Therefore, there is beneficiation as well as good 
recovery of the uranium from the ore deposit. 
The uranium enriched leach solution is withdrawn from the ore deposit. The 
amount of dissolved uranium may be from 100 to 1000 ppm in the solution, 
depending on the richness of the ore, the length of time of contact of the 
leach solution with the ore and porosity of the deposit, as well as other 
factors. 
A convenient method for carrying out the leaching of the ore is to dispose 
a patterned array of injection wells about a roll front ore deposit, the 
well walls having perforated sections at the ore zone between the upper 
and lower mudstone strata shown in FIG. 1, so that leach solution can be 
forced from the well into the ore zone. Spaced a suitable distance from 
the injection wells have several withdrawal wells also having perforated 
sections in their walls in the ore zone so that leach solutions enriched 
in uranium by reason of their passage through the ore can enter and be 
pumped out to a surface treatment plant for extracting and recovering the 
uranium therefrom. A hydraulic gradient is maintained between the 
injection and withdrawal wells so the leach solution will flow from the 
former to the latter. 
Referring to FIG. 3 of the drawing, there is shown a flow sheet 
illustrating the overall general practice of the present invention. The 
leach injection solution comprises from 0.5 to 5 grams per liter of 
ammonium bicarbonate and from about 0.1 to 3 grams per liter of hydrogen 
peroxide, ordinarily added as aqueous hydrogen peroxide, and from 3 to 100 
ppm of iron added as FeNH.sub.4 EDTA wherein the EDTA:Fe ratio is at least 
1.25:1, and sufficient ammonia is added to bring the pH of the solution to 
from about 7.4 to 9.5. The hydrogen peroxide is preferably added to the 
leaching solution immediately before it is delivered into the injection 
well leading to the underground ore deposit in order that the hydrogen 
peroxide does not decompose prematurely. The hydrogen peroxide could even 
be added to the leach solution in the well casing, making sure that it is 
mixed in before the leach solution enters the ore body. The leach solution 
in the injection well is usually under a pressure of from 50 to 250 psi. 
The pressure depends in part on the permeabiity of the sandstone and in 
part on the distance of the injection well from one or more withdrawal 
wells. It should be understood that injection of leach solution may be 
carried out in one or more injection wells either simultaneously or 
serially. 
It has been found that good results are obtained when the injection wells 
are preferably disposed about one or more centrally located withdrawal 
wells which may be spaced to provide a distance of from about 20 to 100 
feet between an injection well to the nearest withdrawal well. In some 
cases the injection well may be located advantageously on the upper side 
of the natural hydraulic gradient with respect to the withdrawal wells. 
Carefully located perforations are provided in the well casing to permit 
leach solution to flow directly only into the ore zone. The leaching 
solution is passing through the ore body will oxidize the ferrous iron to 
ferric iron and the tetravalent uranium present is converted to the 
hexavalent state. The ammonium bicarbonate in the leach solution reacts 
with and readily dissolves the hexavalent uranium in the form of uranyl 
dicarbonate complex. Very little iron from the ore dissolves in the leach 
solution. As the leaching solution contacts the uranium in the sandstone 
and oxidizes and then dissolves the resulting hexavalent uranium, it 
exposes any previously shielded or underlying tetravalent uranium which in 
turn is oxidized and then dissolved. After passing through the sandstone 
or other formation the enriched leach solution passes through perforations 
into the recovery or withdrawal well or wells. Usually a pump will be 
placed in the bottom of the recovery well and the water head in the 
recovery well is maintained at a low level in the well so that there is a 
low hydraulic pressure in the formation adjacent the withdrawal well. 
Consequently, a hydraulic gradient extends from the injection well to the 
withdrawal well thereby causing the leach solution to flow or percolate 
through the sandstone formation toward the withdrawal well. If desired, 
the withdrawal well may be capped and a pump at the bottom energized to 
draw a vacuum with respect to the surrounding ore deposit so that leach 
solution is drawn more strongly to the withdrawal well. 
Tests have indicated that the dilute ammonium bicarbonate solution 
dissolves hexavalent uranium in substantial preference to the other 
mineral values which are also soluble to some extent in this leach 
solution. Thus, assuming an arbitrary ratio of the uranium to the other 
elements in the ore deposit is 1000 to 100, in the enriched leach solution 
the ratio of the uranium to the other elements may be of the order of 1000 
to 5 to 10. Thus, a roughly 10 to 20 fold improvement in the proportion of 
the recovered uranium with respect to the other elements is obtained by 
use of the dilute alkaline ammonium bicarbonate leach solution. 
Thus, in one case, where the ore body had 1269 ppm U (calculated as U.sub.3 
O.sub.8), vanadium 106 ppm, arsenic 12 ppm, molybdenum 8 ppm and selenium 
3 ppm, the ratio of uranium to vanadium was 12, the ratio of uranium to 
arsenic was 106, the ratio of uranium to selenium was 403 and uranium to 
molybdenum was 159. After passing an aqueous leaching solution containing 
0.95 gr. per liter of ammonium bicarbonate and 2.2 grams per liter of 
hydrogen peroxide, the solution exhibited a uranium to molybdenum ratio of 
6800, a uranium to arsenic ratio of 5667, a uranium to selenium ratio of 
531 and a uranium to molybdenum ratio of 59. A later test of the leach 
solution from this well showed that it now had a uranium to molybdenum 
ratio of 259, while the uranium to selenium ratio was 15,436. Apart from 
the iron chelate very little extraneous iron is found in the leach 
solution. Consequently, the selectivity of the dilute ammonium bicarbonate 
leach solutions of this invention for uranium as compared to other 
elements is excellent. 
As shown schematically in FIG. 3, the enriched leach solution pumped from 
the withdrawal well is then passed to an ion exchange column comprising a 
strong base anion material such as a granular resin ion exchange material. 
In the ion exchange column, the uranium is preferentially extracted from 
the enriched leach solution with only a small proportion of the other 
elements being extracted. The ion exchange material is caused to progress 
countercurrently to the flow of leach solution so that the solution coming 
directly from the withdrawal well contacts ion exchange material which has 
picked up uranium from an earlier flow of leach solution therethrough, and 
as the leach solution traverses the column of the ion exchange material it 
meets ion exchange material which has absorbed less and less uranium and 
accordingly it will extract more and more of the uranium therefrom, until 
nearly depleted leach solution contacts relatively fresh ion exchange 
material thereby effecting the maximum uranium recovery. 
The basic anion exchange material, for example, a 16 to 20 mesh ion 
exchange resin such as tertiary amine reacted chloromethyl-styrene-divinyl 
benzene resin (as described in Chemical Engineering for Mar. 18, 1963 on 
pages 166 and 167), when it has taken up nearly its maximum amount of 
uranium is removed in increments from the bottom of the ion exchange 
column and treated with an eluting solution. An aqueous eluant is applied 
countercurrently to the so-removed uranium charged basic anion exchange 
material to strip therefrom the uranium as ammonium uranyl dicarbonate and 
the stripped and rejuvenated ion exchange resin is then returned to the 
top of the ion exchange column to recover additional uranium. The 
concentration of uranium in the eluate produced by treatment of the basic 
anion exchange material may be from 5 to 18 grams per liter of uranium, 
computed as U.sub.3 O.sub.8. A number of different continuous 
counter-current ion exchange contactors and eluant recovery systems may be 
employed. Examples of suitable systems are taught in the January 1969 
issue of "British Chemical Engineering," pages 41 to 46 in an article by 
M. J. Slater entitled "A Review of Continuous Counter-Current Contactors 
for Liquids and Particulate Solids." 
The uraniferous aqueous eluate is acidified and then ammonia is added to 
bring it to a pH of about 7 to 8 to precipitate ammonium diuranate (ADU) 
of a high purity. Ordinarily, the purity of the ADU precipitate after 
washing is such that the impurities therein will not exceed about 1%. 
The barren leach solution ordinarily contains dissolved calcium salts. It 
is desirable to remove these calcium ions prior to refortification of the 
leach solution. To accomplish this, the barren leach solution is passed 
through an ion exchange where ammonium is substituted for calcium. The 
treated barren leach solution is then adjusted or fortified with 
additional ammonium bicarbonate, ammonia and hydrogen peroxide and 
reinjected into the ore deposit to extract additional uranium therefrom. 
The ammonium bicarbonate for the leach solution may be prepared by adding 
ammonium bicarbonate to the aqueous leaching solution. However, a 
convenient and, probably the least expensive way of producing the leach 
solution, is by simply passing ammonia and carbon dioxide gases in the 
required proportions directly into the water where they react in situ into 
ammonium bicarbonate. At the same time, a slight excess of ammonia is 
added to bring the pH to the desired value of about 7.4 to 9. Particularly 
good results are had when the pH of the leach solution is about 7.7 to 
8.5. The following Table of equations comprises the basic reactions of the 
process. 
__________________________________________________________________________ 
TABLE PROCESS CHEMISTRY 
__________________________________________________________________________ 
Leaching 
UO.sub.2 (mineral) + H.sub.2 O.sub.2 
.dwnarw. 
UO.sub.3 (oxidized mineral) + H.sub.2 O 
+ 
NH.sub.4 HCO.sub.3 excess 
.dwnarw. 
(NH.sub.4).sub.2 UO.sub.2 (CO.sub.3).sub.2 + NH.sub.4 HCO.sub.3 
excess + H.sub.2 O Pregnant leach solution 
Uranium 
(NH.sub.4).sub.2 UO.sub.2 (CO.sub.3).sub.2 + NH.sub.4 HCO.sub.3 
excess Pregnant leach solution 
Extraction 
by Resin Ion 
+ 
Exchange 
RCl (resin chloride) 
.dwnarw. 
R.sub.2 UO.sub.2 (CO.sub.3).sub.2 . R.sub.2 CO.sub.3 . RHCO.sub.3 
. RCl Loaded Resin 
.dwnarw. 
NH.sub.4 Cl + . NH.sub.4 HCO.sub.3 excess 
Barren leach solution for recycle 
Resin R.sub.2 UO.sub.2 (CO.sub.3).sub.2 . R.sub.2 CO.sub.3 . RHCO.sub.3 
. RCl Loaded Resin 
Elution 
+ 
NH.sub.4 Cl + NH.sub.4 HCO.sub.3 
Eluant 
.dwnarw. 
RCl Stripped resin 
.dwnarw. 
(NH.sub.4).sub.2 UO.sub.2 (CO.sub.3).sub.2 + NH.sub.4 Cl + 
NH.sub.4 HCO.sub.3 Pregnant eluate 
Acidification 
(NH.sub.4).sub.2 UO.sub.2 (CO.sub.3).sub.2 + NH.sub.4 Cl + 
NH.sub.4 HCO.sub.3 Pregnant eluate 
+ 
H.sub.2 SO.sub.4 
.dwnarw. 
UO.sub.2 SO.sub.4 + CO.sub.2 .uparw. + H.sub.2 O + NH.sub.4 Cl 
Acidified eluate 
(NH.sub.4).sub.2 SO.sub.4 + H.sub.2 SO.sub.4 slight 
pH 2.5 
Precipitation 
UO.sub.2 SO.sub.4 + NH.sub.4 Cl + (NH.sub.4).sub.2 SO.sub.4 + 
H.sub.2 SO.sub.4 slight excess 
Acidified eluate 
+ 
.dwnarw. 
NH.sub.3 
(NH.sub.4).sub.2 U.sub.2 O.sub.7 (solid) (ADU) + (NH.sub.4).sub.2 
SO.sub.4 = NH.sub. 4 Cl Uranium (ADU) 
precipitate slurry 
__________________________________________________________________________ 
The enriched leach solution recovered from the ore deposit may contain 
small amounts of fine suspended calcite and clay particles. It is 
desirable to treat the barren leach solution to remove calcium by 
filtering the leach solution prior to reinjecting it into the well. 
It has been found that occasional plugging of the ore deposit may occur in 
areas adjacent the injection well. Consequently, from time to time when 
the flow of the leach solution has diminished appreciably due to such 
plugging, each injection well may be treated by passing therethrough an 
acid, for example from 10 to 100 gallons of acetic acid. The acid 
dissolves any calcite or other reactive plugging materials so that free 
flow of the leaching solution can again take place. 
LEACH SOLUTION COMPONENTS AND PREATION 
In order to convert the tetravalent uranium to the hexavalent state a 
source of oxygen must be present in the leach solution. The preferred 
source of oxygen is peroxide of hydrogen (H.sub.2 O.sub.2). In some cases 
oxygen gas can be injected along with the aqueous leach solution into the 
ore deposit. While pure oxygen gas is preferable, air can be employed. The 
time of reaction and rate of oxidation of the uranium will be lower with 
the gaseous oxygen than with the H.sub.2 O.sub.2. 
While hydrogen peroxide is available in concentrated solutions of up to 
100%, such high strength solutions are hazardous to handle, in addition to 
being quite costly. Commercially available concentrations of aqueous 
hydrogen peroxide found to be useful in the practice of the present 
invention are those of at least 20% H.sub.2 O.sub.2 concentration and 
preferably from about 30% to 40% hydrogen peroxide. 
Ammonium bicarbonate (NH.sub.4 HCO.sub.3) is commercially available as such 
and has been used in the practice of this invention with good results. 
However, outstanding leaching has been secured when the leach solutions 
were prepared from commercial "ammonium carbonate" which actually 
comprises nearly equal parts of ammonium bicarbonate and ammonium 
carbamate. Data will be presented showing these results. 
The ferric ammonium EDTA may be purchased as a solid material, but the most 
convenient form is an aqueous solution having a 50% solids content by 
weight, wherein about a 2% excess of (NH.sub.4).sub.4 EDTA is present as a 
stabilizer. The iron content of the 50% chelate is nearly 7% by 
weight-70,000 ppm. To prepare for a leaching program extending for 21 days 
during which 25 gallons of leachant will be injected per minute into an 
ore bed, there will be needed 108 gallons of the 50% ferric ammonium EDTA 
solution to provide 10 ppm of iron in the chelate catalyst therein. 
Trivalent iron forms a complex with ethylene-diamine tetraacetic acid, of 
the formula Fe(EDTA).sup.-1. The ammonium salt of this complex is very 
stable and does not precipitate in alkali up to a pH of 12. 
The anionic base exchange material will pick up some of the iron chelate 
catalyst. Thus, Ionac A-590 resin becomes saturated with about 0.3 pounds 
of the iron catalyst per cubic foot of resin. This same resin will 
saturate with uranyl carbonate complex at a loading of about 2.3 pounds 
per cubic foot. The ferric ammonium EDTA will build up in the eluant 
solution and from time to time a portion may have to be discarded and 
replaced with fresh solution. 
The vast majority of the ferric ammonium EDTA in the leach solution will 
remain in the ore bed. In some cases, it may be practical to cease adding 
the iron chelate catalyst to the leach solution after a period of several 
days circulation of the full leach solution into the ore bed, and simply 
to recirculate leach solution to which only ammonium bicarbonate and 
H.sub.2 O.sub.2 have been added to the stripped leach solution. 
Tests have clearly established that the function of the ferric ammonium 
EDTA chelate having an EDTA:Fe ratio of at least 1.25:1, is catalytic in 
nature, and its presence in the leach solution greatly improves the rate 
at which the tetravalent uranium is oxidized to the hexavalent state and 
thus dissolves at a greater rate in the alkaline leach solution. The 
ferric component of this chelate is critical. These facts are evidenced 
from tests whose data are plotted as the several curves in FIG. 4. The 
tests were conducted by agitating 1.00 gram of the powdered UO.sub.2 in 
100 milliliters of a 0.15 molar ammonium bicarbonate solution (about 6 
grams per liter of NH.sub.4 HCO.sub.3 introduced as ammonium 
bicarbonate-ammonium carbamate mixture), and oxygen gas was bubbled 
therethrough and the results are shown in curve 1. The addition of 
(NH.sub.4).sub.4 EDTA only to the leach solution of curve 1 resulted in a 
lower rate of dissolution of the uranium as is evident in curve 2. Curves 
3 and 4 show a great increase in the rate of dissolution of the uranium 
when two different amounts of ferric ammonium EDTA were added to the leach 
solution of curve 1. The introduction of 200 milligrams of Fe as ferric 
ammonium EDTA results in a slight lowering of the dissolution rate of the 
UO.sub.2 as compared to the solution with 100 milligrams of Fe in the 
catalyst; however, this difference may be due to experimental error and 
test variables. 
The criticality of the EDTA:Fe was established in a series of tests made 
employing otherwise identical leach solutions except that the ferric 
ammonium EDTA used had different EDTA:Fe ratios. The curves of FIG. 5 were 
plotted from tests wherein the solutions were similar to those used for 
carrying out that of curve 3 of FIG. 4, and the tests were also similarly 
conducted. As shown by the curves of FIG. 5 ferric ammonium EDTA with 
EDTA:Fe ratio of 1:1 has no appreciable catalytic effect. Iron chelates 
with an EDTA:Fe ratio of 1.5:1 show an enhanced solubilization of 
UO.sub.2, while further enhanced catalysis occurs when the EDTA:Fe ratio 
goes to 2:1 and 3:1. No increased benefit occurs when the EDTA:Fe ratio 
exceeds 2:1. Other tests were made with the ferric chelate catalysts at 
ratios between 1:1 and 1.5:1, and it was found that beneficial catalytic 
results were obtained when the EDTA:Fe ratio was about 1.25 and higher. 
From a number of tests it was found that for each ton of uranium ore 
containing about 0.1% of uranium (2 pounds of uranium per ton or ore), 
sufficient ferric ammonium EDTA should be present in the applied leach 
solution to provide 0.2 pound of iron. In general, from 0.05 to 0.5 pound 
of the ferric ammonium EDTA should be supplied for each ton of uraniferous 
ore containing from about 0.2% to 0.02% of U.sub.3 O.sub.8 by weight. In 
practice, a slightly greater amount of the chelate may be necessary to 
replace losses from spillage, escape into rock fissures and aquifers. 
The agitation leaching tests used in obtaining the data of FIGS. 4 and 5 
constitute highly accelerated conditions as compared to the far slower 
flows of the leach solutions which would occur in underground percolation 
leaching of ore. Therefore column leach tests were conducted to simulate a 
little more closely actual field conditions. Six columns were each packed 
with 90 l grams (dry basis) of uranium ore from the Irrigaray field in 
Wyoming. The ore samples were analyzed and average 0.086% uranium, 1.5% 
iron, 0.022% vanadium, 0.023% copper and 0.0012% molybdenum. Six different 
leach solutions were prepared for circulation through the columns as 
follows: 
A. 10 ppm Fe (in NH.sub.4 FeEDTA.sub.1.5), 50% excess (based on the 
NH.sub.4 FeEDTA.sub.1.5) of (NH.sub.4).sub.4 EDTA, 0.5 g/l H.sub.2 
O.sub.2, 1 g/l NH.sub.4 HCO.sub.3 -pH 7.97 
B. 10 ppm Fe (in NH.sub.4 FeEDTA), 50% excess (NH.sub.4).sub.4 EDTA, 
O.sub.2 gas, 1 g/l NH.sub.4 HCO.sub.3 -pH 7.97 
C. 0.5 g/l H.sub.2 O.sub.2, 1 g/l NH.sub.4 HCO.sub.3 -pH 7.97 
D. 10 ppm Fe (in NH.sub.4 FeEDTA.sub.1.5), 0.5 g/l H.sub.2 O.sub.2, 1 g/l 
NH.sub.4 HCO.sub.3 -pH 7.97 
E. 10 ppm Fe (in NH.sub.4 FeEDTA.sub.1.5), 0.5 g/l H.sub.2 O.sub.2, 1 g/l 
NH.sub.4 HCO.sub.3 and ammonium carbamate-pH 8.94 
F. 0.5 g/l H.sub.2 O.sub.2, 1 g/l NH.sub.4 HCO.sub.3 and ammonium 
carbamate-pH 8.93 
In the solutions of E and F, the ammonium carbonate and the ammonium 
carbamate were in nearly equal proportions. These leach solutions were 
passed through the columns for 70 hours, and the solutions were analyzed 
from time to time to determine the percentage of uranium extracted from 
the ore by each solution. 
As might be expected, the lowest extraction rate was by solution B which 
used oxygen instead of H.sub.2 O.sub.2. The next lowest rate was by 
solution C which had no chelate catalyst. Both solutions A and B showed a 
blue coloration due to the copper extracted from the ore which is due to 
the presence of the excess (NH.sub.4).sub.4 EDTA. Solution D gave the best 
overall rate of the solutions A to D. 
The most dramatic extraction rates were exhibited by leach solution E which 
incorporated the iron chelate catalyst and the ammonium bicarbonate and 
ammonium carbamate mixture, along with the peroxide. Equally important 
with the greatly increased rate of dissolution of the uranium is the fact 
that over 90% of the uranium was extracted by solution E in less than 50 
hours, while solution F extracted nearly 75% of the uranium in the same 
period. 
Solutions D and E contained about 2% of (NH.sub.4).sub.4 EDTA based on the 
weight of the iron chelate, but no blue color was evident after the ore 
was contacted for the entire test time. Consequently it was clear that no 
copper was extracted, at least not to any significant extent. The results 
of these tests are plotted in FIG. 6. 
Referring to FIG. 7, there is schematically illustrated an arrangement of 
apparatus and associated wells for extracting uranium from an underground 
ore deposit in practicing the invention. In a tank 10 there is a supply of 
leach solution which comprises from 0.3 to 5 grams per liter of ammonium 
bicarbonate, and preferably a mixture of ammonium bicarbonate and ammonium 
carbamate, and ferric ammonium EDTA wherein the EDTA:Fe ratio is at least 
about 1.25:1, in an amount to provide from about 3 to 100 ppm of iron, at 
a pH of from 7.4 to 9.5. From tank 10 the leach solution flows through 
pipe 11 to a pump 14 which pumps the solution at a pressure of from about 
50 to 250 psi into a line 12 leading in due course to one or more 
injection wells. The leach solution under pressure is first conveyed by 
line 12 to a filter 16 where any fine clay, calcite and other solid 
particles are filtered out, and thence to a hydrogen peroxide injector 18. 
Aqueous hydrogen peroxide, for example, 30% H.sub.2 O.sub.2, in a storage 
tank 20 passes by pipe 22 to a pump 24 which pressurizes it into the 
injector unit 18 where measured proportions of the hydrogen peroxide are 
injected into the leach solution at a rate to provide from about 0.1 to 3 
grams of H.sub.2 O.sub.2 per liter of each solution. The pipe 12 is 
connected to each of injection wells 26, 30, 32, 34, 36, and 38 by branch 
conduits 25, 27, 29, 31, 33, 35, and 37, respectively, each of the latter 
having valves therein to enable control of flow of the pressurized leach 
solution into one or more wells at any selected time or sequence. 
As is known in the art, the wells comprise a casing which penetrates into 
or through at least one uranium bearing ore stratum. Perforations and/or 
screens are present in the portions of the well casing disposed in the 
uranium bearing stratum to permit pressurized leach solutions to penetrate 
into and through the uranium bearing ore. Normally efforts are made to 
minimize passage of the leach solution to any other stratum by employing 
suitable sealing means between the casing and the well bore. The uranium 
bearing ore stratum, such as the roll front deposits described above 
comprise sandstone, sand or other permeable formations through which the 
leach solution passes readily. The ore bearing formation is usually 
characterized by the presence of an underlying less fluid pervious shale 
or clay layer, and an impervious cap layer of shale, mudstone, or clay so 
that the leach solution flows laterally from the well casing perforations 
into the uranium ore body with very little vertical flow into formations 
containing little or no uranium. The leach solution spreads laterally from 
the injection wells, following the hydraulic gradient, and any channels or 
fissures in the ore body. The hydrogen peroxide in the leach solution 
assisted by the iron chelate catalyst, upon contacting any tetravalent 
uranium converts it into hexavalent uranium which last is readily 
dissolved by the ammonium bicarbonate leach solution. If the uranium is 
present as large particles or as a modular mass or body of appreciable 
size, the uranium is progressively converted, from the exterior surface 
inwardly, to the hexavalent state and each converted surface layer is 
dissolved as successive quantities of leach solution pass and come into 
contact with it. It appears that while much of the iron is converted to 
the ferric state, very little is dissolved in the leach solution, so that 
nearly all the iron remains underground. Similarly, only small amounts of 
molybdenum, vanadium, copper, arsenic and selenium are dissolved in the 
relatively dilute leach solution. By comparison, a high proportion of the 
uranium in the ore deposit contacted by the leach solution is dissolved 
with the passage of sufficient leach solution through the ore body. 
A monitor well 40 is also drilled in order to enable pH, hydrostatic 
pressure, temperature, solution density and other factors to be determined 
by disposing suitable instruments therein. 
In order to withdraw the pregnant leach solution rich in uranium, one or 
more withdrawal wells are disposed in a spaced configuration with respect 
to the adjacent injection wells. A suitable configuration used with 
success is to dispose an encircling array of injection wells about a 
smaller number of withdrawal wells. For example, in one case, seven 
injection wells were spaced to surround three withdrawal wells. 
If there is a strong hydraulic gradient and elongated fissures or the like 
in one direction, the withdrawal wells may be placed so that they are at 
the lower end of the hydraulic gradient in order that the leach solution 
will naturally flow in their direction and be intercepted. 
Disposed within withdrawal wells 50, 52, and 54 are valved connecting 
conduits 51, 53, and 55 respectively, connected by a pump 56 to a pipe 58 
for carrying pregnant solution to a uranium recovery system. While pump 56 
is shown external of the wells, a separate submerged pump disposed in each 
well has been used in each well with good results. Ordinarily, an end of 
each conduit 51, 53, and 55 is disposed at the lower end of the well with 
which it is associated. The withdrawal well comprises a casing with a 
perforated portion with suitable screens to permit leach solution to flow 
into it without allowing sand and other solids to come through. The pump 
56 can be operated to keep the leach solution at any selected level in the 
well so as to maintain a desired hydraulic head therein so that flow of 
each solution therein is established and controlled. Each well can be 
capped and a vacuum applied by operation of the pump 56 so that leach 
solution is attracted more strongly to the withdrawal well. 
The uranium enriched leach solution is transported by pipe 58 to an 
extraction column 60 where it is introduced at the bottom of a bed of, for 
example, 12 to 20 mesh strong base anion exchange resin which moves 
countercurrent to the flow of the leach solution and strips the uranium 
from the leach solution, with however, little of the contaminating metals 
and elements in the leach solution being stripped. 
The depleted or barren leach solution passes from the top of column 60 via 
a pipe 62 to a make-up tank 64. There a concentrated aqueous solution of 
ammonium bicarbonate, and preferably the mixture of ammonium 
bicarbonate-ammonium carbamate previously described, in tank 66 with 
sufficient NH.sub.3 from a storage tank 70 admitted through valved line 
68, is conveyed by pipe 72 by operation of a pump 73 to bring the barren 
leach solution in tank 64 to the desired proportions of from about 0.3 to 
5 grams per liter, and the pH to a value of from 7.4 to 9.5 so that it can 
be recycled. Pump 73 is connected by a valved pipe 69 to a storage tank 71 
containing NH.sub.4 FeEDTA solution with about 2% of (NH.sub.4).sub.4 
EDTA, so that by operation of the valves in pipe 69 the pump 73 can 
introduce the necessary amount of iron chelate solution into the solution 
in make-up tank 64 to bring it up to the desired catalyst content. The 
properly adjusted leach solution in tank 64 is conveyed by pipe 75 to a 
pool reservoir 74 and when needed the leach solution is then carried by 
pipe 76 to the tank 10 from where it is injected into the wells 26-36. It 
will be understood that two or more make-up tanks 64 may be alternately 
filled with barren leach solution while a full make-up tank is being 
adjusted to full strength leaching condition. 
The uranium loaded base anion exchange resin in column 60 is preferably 
removed, either in increments or continuously, from the bottom of the 
column and transferred by a line 82 to an elution column 80 where it is 
treated with strong aqueous solution of ammonium chloride, for example 1.5 
molar, with a small concentration of ammonium bicarbonate, about 0.1 
molar, entering through pipe 86. The uranium is thus extracted by this 
solution from the ion exchange resin to provide a uranium rich eluate. The 
stripped and regenerated ion exchange resin is returned by line 84 to the 
top of column 60 where it again descends and progressively strips uranium 
from the leach solution passing upwardly. 
The eluate with a high content of uranium, for instance, for 10 to 20 grams 
per liter, is carried by pipe 88 from column 80 to either one of two 
ammonium diuranate (ADU) precipitating tanks 90 and 94. When one of the 
tanks 90 or 94 is filled with the eluate, the eluate is then conveyed to 
the other tank. Into the eluate filled tank 90 or 94, a measured amount of 
acid from a supply tank 91 containing HCl, for example, is added to the 
eluate, with suitable agitation and stirring. Then ammonia is added from 
supply tank 70 via pipe 95 to cause the solution to reach a pH of about 7, 
whereupon ADU precipitates. Upon letting the ADU precipitate settle, the 
supernatant liquid is reconveyed by conduit 99 to the eluant tower 80, 
while the ADU slurry at the bottom is pumped to an ADU storage reservoir 
98 through conduit 96. The final ADU product contains only small amounts 
of residual impurities. 
By employing the leach solutions of the present invention, not only is 
there a more rapid recovery of the uranium in an ore deposit, but just as 
important, there will be a more complete recovery of all the uranium in 
the ore deposit, approaching 100% in many cases, all in relatively short 
periods of time. 
Tests on a large ore deposit in Wyoming have begun and preliminary results 
indicate an enhanced recovery of uranium from the ore deposit using the 
iron chelate catalysts of this invention incorporated in the ammonium 
bicarbonate solutions. 
While the leaching process described herein will give excellent results in 
uranium bearing roll front ore deposits, the solutions can be applied to 
other types of ore deposits where the ore body is reasonably permeable to 
the flow of the aqueous leach solutions. In many cases, dense or poorly 
permeable ore bodies can be loosened or fractured by applying controlled 
explosive charges to shatter or to open up the rock structure. The process 
can be applied to mined ore which can be treated by placing the ore in 
large vats or elongated reservoirs and the leach solutions caused to 
traverse the ore and thus dissolve out the uranium. Typical ores that can 
be so processed are coffinite, pitch-blended and uraninite wherein the 
tetravalent uranium is rendered soluble in the leaching solution by 
oxidation to the hexavalent state, and the uranium reacts with the 
bicarbonate component to form the stable uranyltricarbonate complex anion- 
UO.sub.2 (CO.sub.3).sub.3.sup.-4, which is quite soluble in the leach 
solution.