Composition and method for fixing atomic waste and disposal

Sept. 27, 1966

D’ARCY A. SHOCK ETAL

3,274,784

COMPOSITION AND METHOD FOR FIXING ATOMIC WASTE AND DISPOSAL·

Filed Feb. 24, 1961

Sheets-Sheet 1

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A1ISOOSIA

PRESTON L. GANT D'ARCY A. SHOCK JOHN D. SUDBURY

INVENTORS.

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Sept. 27, 1966 d’arcy a, shock etal 3,274,784

COMPOSITION AND METHOD FOR FIXING ATOMIC WASTE AND DISPOSAL

Filed Feb. 24, 1961 ² Sheets-Sheet 3

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FIG. 3

PRESTON L. GANT D'ARCY A. SHOCK JOHN D. SUDBURY INVENTORS.

BY attorney

3,274,784

Patented Sept. 27, 1966

United States Patent Office

3,274,784

COMPOSITION AND METHOD FOR FIXING ATOMIC WASTE AND DISPOSAL

D’Arcy A. Shock, John D. Sudbury, and Preston L. Gant, Ponca City, Okla., assignors to Continental Oil Company, Ponca City, Okla., a corporation of Delaware

Filed Feb. 24, 1961, Ser. No. 88,246

Claims. (Cl. 61—-36)

This application is a continuation-in-part of Serial No. 784,385, filed December 31, 1958, now abandoned.

This invention is directed to composition and method for disposal of radioactive waste. In one aspect, it relates to a method of disposal of radioactive waste solutions and slurries in surface and subsurface earthern reservoirs and the like by means of fixing, in an economical manner, and such that discarded wastes will not present any hazards. In another aspect, it relates to method of disposal of radioactive waste in fractured impermeable formations.

The use of atomic power by the military and the testing of experimental fission reactors already are beginning to cause the accumulation of wastes in substantial volume. An early solution providing for safe and adequate disposal of waste has been sorely needed, although prior to this invention such has failed to be found by the many researchers endeavoring to do so.

The primary difficulty with radioactive wastes, particularly fission product waste, is that they cannot be disposed of by ordinary dilution methods. In order to bring solutions to accepted A.E.C. activity tolerances, the dilution is so great that the method is unfeasible. Retainment until sufficient radioactive decay has taken place to allow dilution seems to be the only solution. In the normal distribution of elements from the fission process, this means a contaminant of approximately 500 to 600 years’ life. As the volume of these wastes increases, it is apparent that the problem of property containing these wastes will become enormous if not already so. For example, one reactor is estimated to consume 13,500 kg. of fuel per year which will yield 3,370 kg. fission product and, after reclaiming the unused fissionable elements, the waste products will amount to about 16,000,000 gallons per year.

The radioactive level of the fission products presents a problem in that they require attention to shielding. The amount of shielding required depends, naturally, on the amount and energy distribution of the radioactive elements. The fission products vary somewhat with the characteristics of a given reactor; however they mainly consist of a large number of short-lived energetic elements and a few long-life members. Thus, the shielding problem becomes less with time. As a range, the fission products in a fuel rod at 30 days’ cooling requires 5½ feet of concrete; in 6 years the shielding required would be less than 1 foot. From that time on, however, the shielding thickness changes very little due to the exponential character of radioactive decay and the presence of some long half-life energetic isotopes. Regardless of the exact amount of shielding required due to preaging, it is obvious that shielding is a problem and a prerequisite. An operation with maximum shielding is not only advantageous but required.

The release of radiation energy due to radioactive decay gives rise to another problem, namely, heat evolution. The amount of heat evolved is related directly to radioactivity of the waste material. The actual temperature attained is dependent upon the rate that heat is transferred out of the system. Since efficient shielding systems are good heat insulating systems, heat removal means must be employed. Proposed methods wherein the radio2 active wastes have been added to aquifers to improve heat dissipation have been only partially successful.

Prior art methods for the disposal of such wastes which have been suggested include the injection of radio5 active waste into depleted oil reservoirs or other subterranean porous formations, underground caverns, abandoned mines, salt domes, aquifers, clays having ion exchange ability, and into cribs and the like; storage in steel-lined concrete tanks; and injection into the sea.

]θ None have been entirely successful.

Specifically when such materials are injected into subterranean formations, there is the ever present danger of subsurface movement or migration of the radioactively hot waste, a potential health and economic hazard.

In summary, any real solution to the problem of disposal of radioactive wastes must necessarily cope satisfactorily with extremely large volumes of substances having ionic compositions of relatively high concentration and often containing substantial amounts of solids, high in20 tensity radiation, high levels of heat energy, and long retention times.

It is a principal object of the present invention to provide a method of disposal of fission products that will overcome the shortcomings of prior art in a safe and 25 economical manner.

Another object is to provide an expeditious and facilitory manner of disposing of radioactive wastes.

These and other objects are accomplished by the present invention which will be fully understood from 30 the ensuing discussion .

The invention broadly comprises solidifiable compositions of matter comprising radioactive wastes in the form of a solution or slurry in water admixed with clay minerals, lime and caustic, said clay minerals, lime and caustic 35 being present in proportions to provide a solid mass on standing.

In one aspect, the invention comprises mixing a radioactive waste solution or slurry with clay minerals to form a slurry similar to drilling mud, converting the slurry to ⁴θ a high lime mud having a pH of above 10 by the addition of lime and caustic. The above procedure will be found sufficient in many cases to attain solidification; however in others it will be found necessary for the gel to be heated to above 500° F. by self-absorption of the heat evolved by the radioactive emission, and/or heat from an outside source. It must be pointed out that, generally, the additional heat is for the purpose of shortening the time for solidification not to cause solidification per se.

In another aspect, the invention is directed to the dis_g0 posal of radioactive waste in impermeable rock formations in horizontal fractures provided therein, by injecting a solidifiable material containing radioactive waste into said fracture and thereafter injecting additional solidifiable material free of radioactive waste to seal the radio₅₅ active waste in the impermeable rock formation. In a more specific aspect of this method, injection of a solidifiable material free from radioactive waste into the fracture is commenced, thereafter radioactive waste is introduced to the flowing solidifiable material and finally ₆₀ the introduction of radioactive waste is terminated and flow of the solidifiable material into the fracture is continued for a period of time to seal the radioactive waste in the impermeable rock formation.

Usually the radioactive waste which is to be disposed θ- of is present in the form of a solution. However, solids can be present, varying from minor amounts to substantial amounts, up to as high as 50 percent by volume of the waste mixture. The solids contained in the waste will, of course, affect to some extent the amount of clay required for sealing and solidification of the composition of this invention. This does not, however, alter the nature of the compositions and the process employing said com

3,274,784 positions; and wastes containing solids (slurries) are treated in the same manner as those wastes which are solutions.

The concentrated radioactive waste solutions and slurries contain fission products such as strontium, cesium, ruthenium, lanthanum, barium and water. The wastes additionally generally contain nonradioactively hot but chemically reactive materials, such as salts and acids. Represenative examples of these are A1(NO₃)₃, HNO₃and H2SO4. Such chemically reactive materials as these, it is believed, prevent the hydration of clays. It has been found, however, when lime and caustic are added, these acid contaminants are neutralized. When sufficient lime and caustic are added to give the solution a pH above about 10, preferably in the range of 10-13 and more preferably from about 10.5 to about 12, solidification of the gel will occur or can be made to occur in a suitable time by heat.

By a method of this invention, it is possible to inject the pretreated radioactive waste into abandoned, substantially depleted oil reservoirs, subterranean caverns, and the like as a solution or slurry. After the radioactive waste has been placed in such disposal locations, they are then solidified in place, and the fissionable materials are entrapped or fixed by adsorption on the clay. In this form, any movement or migration of the hazardous elements is prevented.

By means of this invention, radioactive waste is fixed in place. This prevents escape, which is possible when radioactive waste is in an untreated fluid state. While heat evolved in the present case will cause some formations into which it is injected to be fused, no deleterious results are incurred as with untreated waste. For instance when a solution or slurry treated in accordance to the present invention is injected into a salt dome, the salt dome upon becoming fused will flow to other strata; however, upon reaching these other strata, which will be cooler by comparison, the fused salt will resolidify in place. Any cavity formed around the waste by said fusion and flow will have no effect upon the waste, since it will have long before assumed a rigid structure. Any contaminating radiation carried by any fused salt will soon be dissipated harmlessly to surrounding strata, and the fused salt will not be able to flow a great distance anyway before it is cooled sufficiently to resolidify. However, since originally some flow of the waste and eventually of some formations is possible, it will be readily recognized that injection should never be made into formations in the near vicinity of oil, water, and other mineral deposits, the recovery of which may be desirable; nor should such be injected into the near vicinity of water tables and the like. This could result in the pollution of the oil and other valuable minerals as well as the source of drinking water for humans and animals ’ before solidification takes place and even subsequent thereto. The possibility of some flow is not a disadvantage of the present method; nor is the requirement of some discretion in the location of disposal. It surely can be appreciated that with or without limited flow, radiation damage can occur without direct contact; however this invention does reduce substantially the distances from contaminable useful mineral and water sources at which waste can be disposed without contamination resulting directly or indirectly from movement of the waste. Any flow possible with this invention is in terms of feet and over a limited time, rather than flow and migration distances of miles over an indefinite period.

The present invention makes it possible also to recover some of the radioactive materials and energy from such wastes by their being stored in solid state. This may be accomplished by solidifying in surface reservoirs, removal from, and then placement into not too difficult accessible subterranean caverns, abandoned mines, and the like from which it may be recovered at some future date. Such recovery, of course, will not be without its problems, but these can be coped with satisfactorily. Eventual recovery is not an essential feature of this method anyway, but its making such secondary beneficial utilitarian advantages possible does enhance its desirability as a commercial process.

Many variations and modifications of the method of disposal are possible in the pursuit of this invention. For example, as well as injection into caverns and the like, the wastes may be injected into artificial fractures in subterranean formations with concomitant disposal. This will require the use of the fracturing apparatus and procedures different from simple injection into subterranean natural occurring caverns and the like. A particular application of the invention involving waste disposal in artificial fractures will be described in detail hereinafter.

Some variables involved in carrying out certain aspects of the invention have already been pointed out briefly. These and others which are of some importance will be discussed more fully hereinafter. Naturally, it will be impossible to demonstrate all the limits due to their great number. It will be understood, of course, that all the variables are more or less interdependent, and that, when one is arbitrarily fixed, the limits within which others may be varied are somewhat restricted. The more desirable ranges for ordinary applications of our invention will be indicated and can also be ascertained from the specific examples and teachings presented hereinafter. Likewise, for any particular application of our invention, the most desirable conditions can be readily determined by trial by one skilled in the art; such a determination also is facilitated by the discussion of trends of these variables presented herewith. The discussion will also enable one to proceed without difficulty when his conditions vary or he chooses conditions (materials, etc.,) which vary some from those exemplified in detail herein.

The broad variables with respect to solidification of the specific compositions of this invention are time, temperature, pressure (usually minor effect), the nature of the waste, and the nature of the treating ingredients, the disposal location.

Time, of course, is a dependent variable and is effected by the other variables. What is usually desired, however, is a relatively short time for the occurrence of gellation, and the other variables are generally adjusted to achieve this.

Temperature required for solidification may be regulated somewhat and is a factor that is affected by the other variables. Usually the temperature that will be required for solidification can be adjusted by the major factor affecting this, that is, the particular clay or clay mixture employed. This is best explained that, when separate samples of the same waste were treated in substantially the same fashion and with the same materials with the exception that different clays were employed, one of these treated radioactive waste solutions solidified at room temperature while the other required additional heating. Specifically, one clay found to become very viscous and immobile in approximately 48 hours (it solidified upon further standing) at room temperature when utilized in this invention is X ACT CLAY. X ACT CLAY is sold by Magnet Cove Barium Corporation. X ACT CLAY is defined as predominantly a calcium montmorillonite which does not have the high swelling characteristics of sodium montmorillonite. Bentonite was found to require temperatures on the order of 250° C. or above in order for solidification to take place in about 30 days.

More clay content is required for solidification, with the same quantity of waste in the case of natural clays; therefore a waste fixing system utilizing natural clay is capable of taking more solid contained in the waste. It follows that natural clays are used in different ratios than in the case of bentonite. If natural clays are employed, greater amounts are required than with bentonite.

3,274,784

Natural clays are usually preferred, especially where the waste contains solids.

Lime is added to a waste solution mixed with bentonite in a range based on weight of 2-30 percent based on initial weight of waste. The preferred percentages of lime in bentonite treated waste are 4.0-14 percent of the total weight. Lime used in the above proportions will result in the desired intermediate degree of basicity and a viscous flocculated clay when bentonite is employed. It is to be noted that these specific limits are in the case of bentonite and lime only. With other clays, with wastes differing in types of components and differing considerably in percent of components, this range will vary. Less than 2 percent lime may be all that is required sometimes. But as a general rule, the range 2-30 will be found suitable regardless of these factors. The preferred range of lime with natural clays is 3-15 percent. The lime required may be also defined by a general rule of thumb, which we have found and included herein, for added convenience of understanding the quantity to be employed which will generally insure solidification. This rule is that usually sufficient lime is added so as to provide for a total calcium ion concentration of at least Vi and usually not more than 25 percent in the total weight of the solution containing all the components exclusive of pH adjusting caustic. This is not an absolute but is a convenient rule which will be found highly successful. It can be readily appreciated that, where calcium ions are present in the solution from other sources than the lime (e.g., the immediate water supply, etc.), less lime will be required. Generally, though, at least an over-all concentration of Vz percent of calcium ions is required.

The mechanism of the present invention appears to involve hydration. At some point in the addition of lime, the solution usually becomes a very viscous clay flocculate. At the occurrence of such phenomena, it may be found advantageous to thin the flocculate with a standard mud thinner, such as quebracho, tannins, and the like. Several of such mud thinners are well known in the art; therefor listing them all is unnecessary.

After addition of the lime, the pH must then be brought to above about 10 by the addition of caustic. The caustic which may be employed for this purpose is any commercial grade of basic metal hydroxide such as NaOH, KOH, Ca(OH)₂, etc. Naturally, very weak solutions of caustic should not be employed, as this will cause undue dilution. Generally, caustic solutions on the order of 50 to 75 percent are to be preferred, although either weaker or stronger solutions will be found operable. The most preferred caustic is that in pellet or solid form, being essentially purse caustic. Some caustic materials are available only in such form. The particular alkaline agent is of little concern as long as the required pH is obtained. At the pH above 10, the treated radioactive waste may be solidified at temperatures as low as approximately 20° C. or at least by heating. The pH is somewhat critical inasmuch as below a pH of 10 unduly high temperatures are required for solidification regardless of other factors. Above a pH of 13 excessive amounts of caustic are required which increases the cost of disposal.

While injection into subterranean strata is best not made shallower than 50 feet so as to provide sufficient shielding at the surface, it is also best that injection not be made above the water table in any location. Still further, inasmuch as a temperature in excess of 250° F. may be necessary in some cases to bring about solidification in the desired time, injection in such cases is preferably made at sufficient depths to enable this temperature to be attained more rapidly. Increasing the depth is not the sole answer to obtaining a desired elevated temperature. In the absence of such measures, however, additional heat will have to be supplied. The preferred depth of injection will vary initially because of the subsurface temperature desired and additionally with the geographical area of disposal. This is so because, as is well recognized, formation temperatures at the same depth vary in different geographical areas. This will likewise vary somewhat with the particular strata and its arrangement. Such factors are known in many cases; and where they are not known, they may be readily determined in the usual manner known to those in the art and commonplace in these times.

The invention will be found operable over a fairly wide range of limits with respect to the quantity of clays. To begin with bentonite may be mixed with waste in a range of 1-9% by weight of bentonite based on total weight of the waste. More preferably, the range of 2.0-4.0 percent bentonite by weight of the total weight of waste is employed.

In addition to bentonite, as a gelling compound suitable herein, a natural clay or a mixture of natural clays and complex mixtures of natural clays with bentonite may be used. Natural clays contain a mixture of clay minerals such as montmorillonite, kaolin, chlorite, and attapulgus clays. Natural clays are employed in a ratio of 20 to 60 percent by weight of the total waste, and preferably 25-45 percent. Other clays and mixtures will vary between that when bentonite alone is used and natural clay alone is used. The over-all range in percent by weight of the clay, either bentonite or natural clay mixtures, is 1-60 percent, the lower limit being when bentonite alone is used and the upper limit when a natural clay is used.

Because of the differences in the quantity of clay and time involved for solidification, mixtures may be used advantageously to adjust the time of solidification. This increases the flexibility of the solidification time, inasmuch as mud thinners may also be used for this purpose. Solidification is desired in a short time but time is also considered in light of the discussion below. The disadvantages of requiring too long a time has been explained as possibly permitting the radioactive waste-containing mixture to have too much liberty to flow. The disadvantage of the mixture setting up too soon is that it hampers injection. Solidification is usually not desired before injection is complete. The disadvantage of solidification in the injection apparatus is readily recognized; but if injection is to be made such that the treated waste mixture extends radially from the injection well over a distance of a mile or more (or even less) which will involve some time, solidification at the outer extremities may occur before injection is complete and would hamper injection. Solidification of the waste before injection is complete may be permitted, providing the solidification is not within the injection apparatus or well. Solidification at a point beyond this would require special apparatus such as for fracturing in order to continue injection. This is entirely permissible; however injection without such requirements would be preferred, and solidification would best be regulated such that injection is substantially complete first before the waste solidifies.

In adjusting the solidification time as well as for the purpose of imparting some desired properties to the treated waste after solidification, a cement may be added in addition to the ingredients already mentioned. The cements which will be found suitable are those now being used by the petroleum industry for injecting into subsurface wells for varying reasons. Portland cement, to name one, is suitable and illustrative, however the inven·’ tion is not limited thereto. Several of these cements are well known in the art. Additional mud thinners such as those previously mentioned may be used (sometimes it will be required) to inhibit premature setting of the concrete during injection. Mud thinners are commonly used for such purposes.

Pressure has lesser effect than the other variables; but as might be expected, increased pressure has the general effect of speeding solidification. One simple solution to

3,274,784 obtaining increased pressure, where it is sufficiently helpful to warrant increased pressure, is to inject it into subterranean formations at some relatively great depth. For this purpose, injections at 1,000-foot depths may be desired and ΙΟ,ΟΟΟ-foot depths or still deeper may be de- 5 sired, depending on the precise pressure desired to be exerted on the treated waste. Since pressure has such a minor effect, depth is usually ignored on the basis of pressure; and the depth of injections is generally based upon other considerations. jq

Another important variable is the composition of the waste to be disposed of. These vary from reactor to reactor and particularly with the different types and processing of the fuel element from these reactors. Representative examples of the compositions of wastes result- 15 ing from present commercial reactors or reactors which appear destined for commercial acceptance and operation in some number are shown herein below. This information is based on declassified and released information now available to the public. For convenience, methods of 20 simulating these wastes are also included which will enable one to determine the variables in a case varying some from that specifically shown herein. Thus the variables may be better determined for the particular case in small scale lab tests with routine approach using 25 the teachings of this invention.

WASTE FROM PROCESSING OF ALUMINUM-ENRICHED URANIUM ALLOY FUEL ELEMENTS

ΟΛ

The TBP-25 process used in separating and decontaminating enriched uranium from uranium-aluminum alloy fuel elements, such as the fuel of the Materials Testing Reactor (MTR) at the Idaho Chemical processing plant, is as follows: The fuel is dissolved in nitric acid, with mercuric nitrate as catalyst; and the uranium ³⁵is extracted with 5 percent tributyl phosphate in a kerosene-type diluent—both nitric acid and aluminum nitrate are salting agents. The aluminum wastes are not neutralized and are stored in stainless steel tanks. ,θ

The average chemical composition of a typical Idaho Chemical Processing plant (ICPP) process waste solution is:

Volume, gal./g. U-235 -----------------Specific gravity________________________

H+__________________________________

Al³+_________________________________

Hg++-------------------------------NO3--------------------------------SO4—________________________________

Fe³+_________________________________

NH₄+________________________________

0.22 to 0.11.

1.29.

0.5 M.

1.6 M.

0.01 M.

5.5 M.

0.02 M.

0.002 M.

0.05 M.

The radioactivity of the waste is 2,500 to 5,100 curies/ ₅₅gal.; the energy is 44 to 88.6 B.t.u./hr. gal. or 12.5 to 26 watts/ gal.

Simulated first extraction column waste is prepared from the following chemicals:

GO

Chemical

Α1(ΝΟ₃)3·9Η₂Ο__________

FeaCSOrs'XILO--------(NH₄)2SO₄_______________

Hg(NO₃)₂________________

NH4NO3_________________

HNO3____________________

Molecular Wt.

Amount Used per Liter

Final Cone., M

375 400

132 325

63

600 g______________

0.8 g.*____________

2.64 g------------3.25 g------------2.4 g-------------33 ml. of 15.6 M...

1.6

0.002

0.02

0.01

0.03

0.5 *To be added as Fe2(SO₄)3.

The chemicals are dissolved in water and diluted to one liter. 75

WASTE FROM PROCESSING OF ZIRCONIUM URANIUM FUEL ELEMENTS

The Submarine Thermal Reactor (STR) of the Nautilus has a zircaloy (98% Zr-2% Sn) clad zirconiumenriched uranium alloy core. The choice of dissolution reagents is very limited with this corrosion-resistant material, and the elements are currently being dissolved in hydrofluoric acid. The uranium (IV) is oxidized to U(VI) by chromate and extracted with 10 percent tributyl phosphate. During this oxidizing step the Al(HI) complexes the free fluoride ion.

A typical actual composition of first extraction column waste is:

Volume, gal.--------------------------- 2,500.

Specific gravity__________________________ 1.216.

H+___________________________________ 2.14 M.

Al³+__________________________________ 0.75 M.

Zr⁴+___________________________________0.55 M.

Sn++__________________________________0.012 M.

Cr³+__________________________________0.016 M.

NO3-_________________________________ 3.59 M.

F-____________________________________3.00 M.

This first extraction column waste may be simulated from the following chemicals:

<td> Chemical</td><td> Molecular Wt.</td><td> Amount Used per Liter</td><td> Final Cone., M</td>

<td> Zr(OH)₄.12H₂O___________</td><td> 375</td><td> 206 g--------------</td><td> 0. 55</td>

<td> A1(NO₃)₃.9Ξ2Ο___________</td><td> 375</td><td> 273 g--------------</td><td> U. 15</td>

<td> ur(fM 03)3. yii2U__________</td><td> 400</td><td> 6.4 g---------------</td><td> 0. 016</td>

<td> Sn (mossy)_______________</td><td> 119</td><td> 1.43_______________</td><td> 0. 012</td>

<td> HF_______________________</td><td> 20</td><td> 107 ml. of 28 M____</td><td> 3. 0</td>

<td> JtLNUj____________________</td><td> 63</td><td> 83 ml. of 15.6 M____</td><td> 1. 3</td>

The radioactivity of actual waste would be 100 curies/gal.; the energy emitted would be 1.71 B.t.u./hr. gal. or 0.5 watt/gal. (The zirconium hydroxide may be more or less difficult to dissolve, depending upon the temperature used in drying.)

The simulated waste may be prepared in the following manner: Dissolve 50 g. of Zircaloy-2 in 340 ml. of 2.37 M. FIF to provide the zirconium and tin. To this is added the nitric acid and water solutions of the other salts, and the entire solution diluted to one liter volume. The nitric acid and salts should not be added until all the Zircaloy-2 is in solution. The hydrofluoric acid should be added to the Zircaloy slowly. Although warm water will attack magnesium, warm HF will not do so because of the formation of a protective film. The waste is stored in a polyethylene bottle.

Note: Zirconium as a powder or sponge metal has a very great tendency to ignite spontaneously in air at low temperatures. The combustion hazard is greatly increased when 16 percent or less moisture is present. The U.S. Bureau of Mines reported that dry zirconium powder (90% of 6μ size or smaller) can ignite explosively and spontaneously when dispersed in air at room temperature. No hazard exists during the dissolution of zirconium or Zircaloy-2 if hydrofluoric acid is added slowly at first. Care should be taken because of displacement of hydrogen by the metal.

WASTE FROM PROCESSING OF STAINLESS STEEL-UO₂ FUEL ELEMENTS

The Darex process is used in processing uranium fuel elements containing stainless steel, like types 304 and 347, which readily dissolve in dilute aqua regia. The chloride is removed to a level of 30 p.p.m. by distillation before solvent extraction of the uranium with TBP. Examples of fuel elements to be processed by this method are the Army Package Power Reactor (APPR), which has a stainless steel jacket and sintered enriched uranium dioxide-stainless steel core; and the Yankee Atomic Power Reactor, which has a stainless steel jacket and slightly enriched UO₂ core.

3,274,784

A typical composition of an actual extraction column aqueous raffinate is:

Volume, gal./g. U-235 0.5.

Specific gravity 1.28.

H+___________________________________ 2.94 M.5

Ni²+__________________________________ 0.075 M.

Fe³+__________________________________0.68 M.

Cr³+__________________________________0.16 M.

Al³+__________________________________ 0.123 M.

NO₃-_________________________________'6.1 M. 10

Cl^-, p.p.m._____________________________ 30.

Simulated waste may be prepared in the following manner:

Chemical

Molecular Wt.

Amount Used per Liter

Final Cone., M

Ni(NO₃)2. 6I-I2O__________

Fe(NO₃)₃.9H₂O__________

Cr(NO₃)3.9H₂O__________

A1(NO₃)₃.9H2O___________

HC1______________________

IINO3____________________

0.075

0. 68

0.16

0.123

0.001

2.95

The nitrate salts readily dissolve in water. The acid should be diluted, mixed with the solution of salts, and 25 diluted to one liter. The appearance of the waste solution will be dark blue-black because of the chromium nitrate.

As can be seen from the representative examples of waste above, the concentration of important ions such as 30 NO3- and SO₄— ions and the like vary; therefore the quantity of caustic in this invention will accordingly vary in different cases. The strength of the particular caustic will naturally have some bearing on the quantity of caustic necessary. Since, however, the quantity of cans- 35 tic added is determined by pH, this will present no problems. The particular clay mineral will also have some effect with respect to the quantity of lime and caustic but will present no problems with the caustic in view of the more detailed discussion of the pH range suitable. 40

In addition to the concentration of ions varying in the different wastes, it will be noted that the solids content in the waste before treatment varies also and will (as previously stated) affect to some extent the particular gelling mud and quantity of same to be used in different in- 45 stances.

As pointed out previously in one modification of the invention radioactive wastes are disposed of by injection in subterranean formations, through the use of artificial fractures provided in the formation. A more detailed 50 description of how such formations are utilized is presented hereinbelow.

A well is drilled into a subterranean formation. The formation is isolated as by cementing. The formation is perforated for example, by gun perforation. It will 55 be found desirable to isolate this perforated zone with packers in most cases. The fracturing fluid, which may be the treated radioactive waste, is pumped into the well under pressure, thereby building up a hydrostatic pressure. When the hydrostatic pressure exceeds the forma- 60 tion breakdown pressure, the formation will part or fracture. Since the pressure ceases to rise when the formation breakdown pressure is reached, fluid pressure measurements at the surface indicate when that point is reached. Stated somewhat more accurately, the for- 65 mation breakdown pressure may be defined as the pressure at which the increase of the rate of fluid injection into the formation will not materially increase the fluid pressure. After the fracture is filled with the waste liquid, the fracture may be sealed as for example, by 70 cementing. If desired the fracturing, filling the fracture and sealing may be repeated at different levels in the same formation.

A very rough estimate of the pressure in pounds per square inch required for fracturing a formation at any 75 particular depth is numerically equal to Vz the depth in feet of that formation. This pressure varies, however, from place to place, depending upon the depth and the nature of the formations (e.g., densities, etc.), folding of the formations, and the like. A particularly advantageous method of disposal suitable for this invention is to inject into fractures in impermeable formations using procedures according to the invention more fully described in assignee’s copending application, Serial No. 30,988, filed May 23, 1960, now U.S. Patent No. 3,108,439, a continuation-in-part of Serial No. 762,991, filed Sept. 24, 1958, now abandoned. The latter application corresponds to the issued French counterpart Patent No. 1,235,240.

The process of injecting radioactive waste into impermeable formations in accordance with the method of this invention comprises introducing the waste in a solidifiable material into the fracture in the impermeable formation and thereafter sealing the formation by introducing an additional amount of solidifiable material free from radioactive waste. The solidifiable materials which can be employed in this particular method of disposal include various gelling compositions, cements, muds and other materials which are compatible with water, are readily pumpable and are capable of retaining radioactive materials therein upon solidification. The combinations of clay, lime and caustic contemplated in the specific compositions of this invention can be used and also more conventional solidifiable materials, such as, Portland cement and other cements, and the like. While it is usually contemplated using the same solidifiable material. in each step of the aforedescribed process, different materials can be employed if desired, within the scope of the invention.

In the preferred method of disposal in impermeable formations, a flow of solidifiable material is commenced into the fracture. Thereafter, radioactive waste is introduced to the flowing stream, the addition of said waste continuing until the desired amount for disposal has been added. Following this the flow of solidifiable material (free from radioactive waste) is continued for a period of time, whereby the radioactive waste is sealed in the formation. The latter method of operation provides in the formation a solid mass of material free from radioactive waste followed by radioactive waste which is enclosed in a solid mass and thereafter a second body of solid material which is free from radioactive waste. This method of operation provides additional protection against movement of radioactive waste from the formation by assuring that no openings exist at the outer periphery of the fracture and also by assuring adequate sealing material at the inlet to the fracture.

The relative amounts of solidifiable material used in the various steps of the disposal process, that is, the initial flow of solidifiable material, the flow of combined material and radioactive waste and the final addition of solidifiable material free from radioactive waste will vary depending on the size of the impermeable formation, the extent of the fracture, etc.

The spacing of zones of disposal in said copending application (S.N. 30,988) is necessary for heat dissipation, and such spacing will have the same effect when the waste is pretreated according to the present invention. Such heat dissipation, however, will have no deleterious effect upon solidification where heat is required, for example, with the specific compositions of this invention. This is due to the fact that solidification takes place in time at a temperature below that required to fuse the formation.

It may be found desirable to have pilot and check wells located around injection wells at various distances, while the treated waste is injected into subterranean locations. In this manner, the extremity of the injected waste may be determined and will aid in determining when injection

3,274,784 into a specific strata or zone should be stopped due to prevailing conditions.

Unlike assignees’ copending application, treated radioactive wastes in the broad aspect of this invention can be injected into fractures of almost any formation and need not be substantially horizontal, although such would be preferred. In this respect, the waste after being treated, may be disposed of in the same manner as salt water, petroleum wastes, and the like by the petroleum industry such as fracturing and injectng into subterranean strata. In this case as with wastes from the petroleum industry, injection may be made between beds of strata or into a stratum with little regard for the characteristics of the strata. Naturally, some differences exist in that injection should not be made into subterranean reservoirs filled or partially filled with fluids. Preferably, injection of the radioactive waste is carried out in horizontal fractures in an impermeable formation. The obvious advantage of this is, of course, that fracturing in such a formation can provide a reservoir for waste disposal which is impenetrable to the movement of waste and by the procedures of this invention it can be assured that the waste will be retained in the formation indefinitely.

In order to more clearly describe the invention and provide a better understanding thereof, reference is made to the accompanying drawings of which FIGURE 1 is a graphical representation of the quantity of solids necessary for solidification in the separate cases of bentonite and natural clay, a typical natural clay such as used in oil well drilling muds;

FIGURE 2 is a diagrammatic illustration of a process flow suitable for injecting radioactive waste into an impermeable formation; and

FIGURE 3 is a diagrammatic illustration in cross-section of a mixer for mixing solidiflable material and radioactive waste.

Referring particularly to FIGURE 2, pump 8 is a fracturing pump which is connected to an injection well 32 through conduits 10, 14 and 30, containing valves 12 and 16 and a mixer 28. Tank 2, which is adapted to contain a mobile solidiflable material, is connected to conduit 14 through conduit 4 and valve 6. Shielded storage 20, adapted to contain radioactive waste, connects with mixer 28 through pump 22, conduit 24 and valve 26. In the preparation for disposal of radioactive waste, injection well 32 is first provided in an impermeable formation. A suitable fracture or fractures are established in the injection well, utilizing fracturing pump 8 and a suitable fracturing fluid, the fracturing being carried out in the ordinary and conventional manner well known in the art. Any of the usual fracturing fluids can, of course, be employed in this operation. When the underground storage is ready for use, the disposal process is initiated by commencing flow of solidiflable material (e.g., Portland cement) from tank 2 through conduits 4, 14 and 3® and mixer 28 into the injection well 32. After the desired amount of solidiflable material has been introduced to the injection well, the addition thereto of radioactive waste from storage 20 is commenced through pump 22 and conduit 24. The radioactive waste is admixed with the solidifiable material in mixer 28, and the admixture thereafter flows into injection well 32. After the desired amount of radioactive waste has been added, the flow of this material is terminated; and the flow of solidiflable material in the injection well is thereafter continued for a period of time to provide sealing of the radioactive waste in the impermeable formation. This process can be repeated at various levels and utilizing various fractures in the formation. As necessary, additional solidiflable material and radioactive waste can be introduced to the storage vessels through conduits 1 and 18, respectively.

It is desirable that the radioactive waste be thoroughly distributed throughout the solidiflable material. Any conventional mixing device can be utilized for this purpose, however, a device which is particularly applicable is illustrated in FIGURE 3. The mixer of FIGURE 3 comprises a conduit 34 having an area 36 of restricted cross-section, said restriction being provided in a similar manner as a venturi, and a smaller conduit 38 opening within said area of reduced cross-section, preferably on an axis coinciding with the longitudinal axis of conduit 34. In the operation of this mixer, solidiflable material 4® (e.g., a clay) is passed through conduit 34, radioactive waste (solution or slurry) 42 being introduced thereinto through conduit 38 in a direction parallel to the flow of solidiflable material. The operation of the mixer differs from the ordinary aspirator type of mixer in that the pressure in. conduit 38 is maintained higher than the pressure of the solidiflable material in the unrestricted portion of conduit 34. Also the volume of the radioactive waste is maintained less than the volume of the solidiflable material. In addition to providing effective mixing of radioactive waste and solidifiable material, the mixer by proper control of the pressures of the two streams 40 and 42 can be utilized to control the relative amounts of each material introduced to the formation. This mixing valve has the advantage of containing no moving parts and can readily be decontaminated, for example, by flowing solidiflable material or other material through the radioactive waste line 38, while the main stream of solidiflable material is flowing through the mixer.

The following examples are presented in illustration of the invention:

Example 1

Four-hundred fifty-five pounds of a typical waste (from an Al-enriched uranium alloy fuel element) having approximately 25 percent of A1(NO₃)₃ and HNO₃ and approximately one percent of fission products, which are principally strontium, cesium, and the rare earths, is mixed with 105 pounds of low yield clay containing bentonite. Approximately 20 pounds of lime is next added to this solution. Sodium hydroxide (150 pounds of 50 percent NaOH) is then added until a pH of 12 is obtained. At a pH of approximately 8, a flocculation is obtained; and the solution is thinned to a pumping viscosity with a standard mud thinner. The solution, after a pH of 12 is obtained, is pumped into a subterranean location at say approximately 3,000 feet. The treated radioactive waste becomes solid in place in time.

Example 2

A test well was drilled 300 feet deep into a compact shale formation existing at the waste disposal site. The well was drilled 6 in diameter, cased with 3Ά O.D. tubing, and cemented to the surface. Four observation wells were drilled to a 200-foot depth and located in perpendicular distances 200 feet from the injection well. This gave a five-spot pattern with the disposal well as the center well. A number of bench markers were placed in the area to record any ground rise that might result from the well fracture. The observation wells were packed with sand some 100 feet off bottom and then cased and cemented to the surface. The drilling and completion of the wells was performed over a period of time of several weeks before the fracturing experiment began.

The fracturing experiment began by lowering a 2Ά O.D. sand jet tool run on PA tubing and circulating sand for sufficient time to cut a notch in the casing and cement and formation at 290 feet. The length of time was calculated to be sufficient to give an 18-inch deep notch. The sand jet tool was removed with the tubing from the well and two high pressure fracturing pump trucks were connected to the well head. Using water as the hydraulic fluid the well was pressured up and fractured in the conventional manner. The fracture pressure was approximately 2300 pounds, falling to 800 pounds. The pumping rate at this pressure was 7 barrels per minute.

3,274,784

After the fracture was created the disposal was begun. The waste was simulated by mixing 35 curies of radioactive cesium as the chloride in 9 gallons of water. The simulated waste solution was injected into the well head of the disposal well by means of a positive displacement pump. The radioactive material was shielded, as well as the pump and radioactive injection line. Frequent checks were made to demonstrate that the radiation was not over tolerance levels at any point around the installation. The method of placing the radioactive simulated waste was to pump a .12.2 pounds per gallon cement slurry into the fractured well. After the cement injection was stabilized, the simulated waste stream was introduced by the displacement pump into the slurry stream of cement. In this manner the radioactive solution was mixed and dispersed through the cement slurry before it was injected into the shale. Approximately 400 pounds per square inch surface pressure were required for this operation. After 95% of the cesium and 20,000 gallons of cement slurry had been pumped there was a flow detected coming from the up dip observation well. At this point the pumping operation was stopped. The hydraulic pump truck was moved off location, cement allowed to harden, and a core removal program followed to determine the location of the radioactive simulated waste in cement.

In brief, it was found that the simulated waste followed the general structure of the shale which rose a few feet up dip. The fracture was found to extend in a more or less horizontal plane in all directions around the injection well. In no case was there evidence that the cement fractured the shale vertically and tended to rise to the surface. Instead, it followed the contour of the bedding planes of the shale. The test was taken as conclusive evidence that radioactive waste could be placed in a deposit such as the shale without danger of rising to the surface around the well. It also showed that test wells could define the limit of extent of the fracture.

Example 3

The location of the test site was selected to provide a compact shale of sufficient thickness and depth to give a useful test. The injection well was drilled to approximately 1100 feet, cased with 5-inch casing, and cemented to the surface. The well head was equipped for high pressure using both vertical and side arm fittings. The first step was to cut a slot in the casing and shale by running a sand jet tool on 2¹A tubing to approximately the 1000-foot level and pumping sand slurry through the jet for 45 minutes. The sand was displaced and the tubing and jet tool removed.

The well was then fractured with water. The formation broke at approximately 1600 pounds and the water injection was changed to cement injection. When the cement slurry was definitely stabilized the simulated waste was injected by a positive displacement pump through a side connection into the flowing stream of cement. The cement was injected at a rate of 3 barrels per minute for a period of 11 hours until the 50 curies of radio cesium was injected. The well was sealed with a final injection of cement containing no radiotracer, and the pumping equipment was disconnected and this phase of the experiment considered complete.

The process was repeated at the 800-foot level using the sand jet notching technique, then the water fracture, and finally the pumping of cement containing tracer and the sealing operation. There was no evidence of surfacing and it was concluded that this method would satisfactorily retain relatively large volumes of waste fluids and solids. It was also concluded that a single well could be used for a number of waste fluid injections.