US3274784A - Composition and method for fixing atomic waste and disposal - Google Patents

Composition and method for fixing atomic waste and disposal Download PDF

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US3274784A
US3274784A US88246A US8824661A US3274784A US 3274784 A US3274784 A US 3274784A US 88246 A US88246 A US 88246A US 8824661 A US8824661 A US 8824661A US 3274784 A US3274784 A US 3274784A
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waste
radioactive waste
radioactive
disposal
caustic
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D Arcy A Shock
John D Sudbury
Preston L Gant
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ConocoPhillips Co
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Continental Oil Co
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Priority to NL271326D priority Critical patent/NL271326A/xx
Priority to GB36812/59A priority patent/GB926822A/en
Priority to FR814023A priority patent/FR1243673A/fr
Priority to BE597015A priority patent/BE597015Q/fr
Application filed by Continental Oil Co filed Critical Continental Oil Co
Priority to US88246A priority patent/US3274784A/en
Priority to FR871221A priority patent/FR1298142A/fr
Priority to GB40762/61A priority patent/GB940254A/en
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    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21F—PROTECTION AGAINST X-RADIATION, GAMMA RADIATION, CORPUSCULAR RADIATION OR PARTICLE BOMBARDMENT; TREATING RADIOACTIVELY CONTAMINATED MATERIAL; DECONTAMINATION ARRANGEMENTS THEREFOR
    • G21F9/00—Treating radioactively contaminated material; Decontamination arrangements therefor
    • G21F9/04—Treating liquids
    • G21F9/20—Disposal of liquid waste
    • G21F9/24—Disposal of liquid waste by storage in the ground; by storage under water, e.g. in ocean
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21F—PROTECTION AGAINST X-RADIATION, GAMMA RADIATION, CORPUSCULAR RADIATION OR PARTICLE BOMBARDMENT; TREATING RADIOACTIVELY CONTAMINATED MATERIAL; DECONTAMINATION ARRANGEMENTS THEREFOR
    • G21F9/00—Treating radioactively contaminated material; Decontamination arrangements therefor
    • G21F9/04—Treating liquids
    • G21F9/06—Processing
    • G21F9/16—Processing by fixation in stable solid media
    • G21F9/162—Processing by fixation in stable solid media in an inorganic matrix, e.g. clays, zeolites
    • G21F9/165—Cement or cement-like matrix
    • Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
    • Y10—TECHNICAL SUBJECTS COVERED BY FORMER USPC
    • Y10S—TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
    • Y10S55/00—Gas separation
    • Y10S55/09—Radioactive filters

Definitions

  • This invention is directed to composition and method for disposal of radioactive waste.
  • 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.
  • it relates to method of disposal of radioactive Waste in fractured impermeable formations.
  • the primary diificulty with radioactive Wastes is that they cannot be disposed. of by ordinary dilution methods.
  • the dilution 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 properly containing these wastes will become enormous if not already so. For
  • one reactor is estimated to consume 13,500 kg.
  • 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.
  • the fission products in a fuel rod at 30 days cooling requires 5 /2 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 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 trans ferred out of the system. Since efficient shielding systems are good heat insulating systems, heat removal means must be employed. Proposed methods wherein the radio- See 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 include the injection of radioactive 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. None have been entirely successful.
  • 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 intensity radiation, high levels of heat energy, and long I retention times.
  • Another object is to provide an expeditious and facilitory manner of disposing of radioactive wastes.
  • 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 being present in proportions to provide a solid mass on standing.
  • 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 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.
  • the invention is directed to the disposal 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 radioactive waste in the impermeable rock formation.
  • 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 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.
  • the radioactive waste which is to be disposed of is present in the form of a solution.
  • 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 compositions; 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 Al(NO HNO and H 50 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 l-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.
  • 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.
  • 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
  • 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.
  • 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.
  • Time 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 4-8 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.
  • 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 230 percent based on initial weight of waste.
  • the preferred percentages of lime in bentonite treated waste are 4.0l4 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 230 will be found suitable regardless of these factors.
  • the preferred range of lime with natural clays is 315 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 /2 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 /2 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 fiocculate 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.
  • the caustic which may be employed for this purpose is any commercial grade of basic metal hydroxide such as NaOH, KOH, Ca(OH) 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 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.
  • injection into subterranean strata is best not made shallower than 50 feet so as to provide sufiicient 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,
  • 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.
  • bentonite may be mixed with waste in a range of 19% 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.
  • 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 l60 percent, the lower limit being when bentonite alone is used and the upper limit when a natural clay is used.
  • mixtures may be used advantageously to adjust the time of solidification. This increases the flexibility of the solidification time, inasmuch as rnud 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.
  • 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 invention is not limited thereto.
  • 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 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 10,000-fot depths or still deeper may be desired, 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.
  • compositions of the waste to be disposed of 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 resulting 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 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 the teachings of this invention.
  • the TBP-25 process used in separating and decontaminating enriched uranium from uranium aluminum alloy fuel elements is as follows: The fuel is dissolved in nitric acid, with mercuric nitrate as catalyst; and the uranium is extracted with 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.
  • the radioactivity of the waste is 2,500 to 5,100 curies/ gal; 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:
  • the chemicals are dissolved in water and diluted to one liter.
  • 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(III) complexes the free fluoride ion.
  • a typical actual composition of first extraction column waste is:
  • This first extraction column waste may be simulated from the following chemicals:
  • 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 Zircaloy2 in 340 ml. of 2.37 M. HP 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 HP will not do so because of the formation of a protective film. The waste is stored in a polyethylene bottle.
  • 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 US. Bureau of Mines reported that dry zirconium powder of 6 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.
  • 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 ppm. by distillation before solvent extraction of the uranium with TBP.
  • 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 U0 core.
  • APPR Army Package Power Reactor
  • the Yankee Atomic Power Reactor which has a stainless steel jacket and slightly enriched U0 core.
  • a typical composition of an actual extraction column aqueous raflinate is:
  • Simulated waste may be prepared in the following manner:
  • the nitrate salts readily dissolve in water.
  • the acid should be diluted, mixed with the solution of salts, and diluted to one liter.
  • the appearance of the waste solution will be dark blue-black because of the chromium nitrate.
  • the concentration of important ions such as NO; and SO;- 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 caustic 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.
  • radioactive wastes are disposed of by injection in subterranean formations, through the use of artificial fractures provided in the formation. A more detailed 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 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.
  • the hydrostatic pressure exceeds the formation 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 formation 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.
  • the fracture may be sealed as for example, by 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 particular depth is numerically equal to /z 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 accord-ing to the invention more fully described in assignees copending application, Serial No. 30,988, filed May 23, 1960, now US. 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 materialin each step of the aforedescribed process, different materials can be employed if desired, within the scope of the invention.
  • 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.
  • 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 1 1 into a specific strata or zone should be stopped due to prevailing conditions.
  • 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.
  • the waste after being treat ed 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.
  • injection may be made between beds of strata or into a stratum with little regard for the characteristics of the strata.
  • injection should not be made into subterranean reservoirs filled or partially filled with fluids.
  • injection of the radioactive waste is carried out in horizontal fractures in an impermeable formation.
  • 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
  • FIGURE 3 is a diagrammatic illustration in cross-section of a mixer for mixing solidifiable material and radioactive waste.
  • 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 solidifiable material, is connected to conduit 14 through conduit 4 and valve 6.
  • 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, he employed in this operation.
  • the disposal process is initiated by commencing flow of solidifiable material (e.g., Portland cement) from tank 2 through conduits 4, 14 and 30 and mixer 28 into the injection well 32.
  • solidifiable material e.g., Portland cement
  • 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.
  • 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.
  • solidifiable material 40 e. g., a clay
  • radioactive waste (solution or slurry) 42 being introduced thereinto through conduit 38 in a direction parallel to the flow of solidifiable 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 solidifiable material in the unrestricted portion of conduit 34. Also the volume of the radioactive waste is maintained less than the volume of the 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 solidifiable material or other material through the radioactive waste line 38, while the main stream of solidifiable :material is flowing through the mixer.
  • Example 1 Four-hundred fifty-five pounds of a typical waste (from an Al-enriched uranium alloy fuel element) having approximately 25 percent of Al(NO and HNO and approximately one percent of fission products, which are principally strontium, cesium, and the rare earths, is mixed with pounds of low yield clay containing bentonite. Approximately 20 pounds of lime is next added to this solution. Sodium hydroxide 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.
  • a typical waste from an Al-enriched uranium alloy fuel element having approximately 25 percent of Al(NO and HNO and approximately one percent of fission products, which are principally strontium, cesium,
  • 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 3 /2" 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 sevenal weeks before the fracturing experiment began.
  • the fracturing experiment began by lowering a 2 /2" O.D. sand jet tool run on 1 /2" tubing and circulating sand for sufiicient time to cut a notch in the casing and cement and formation at 290 feet. The length of time was calculated to be sufiicient 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.
  • 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 method of placing the radioactive simulated waste was to pump a 12.2 pounds per gallon cement slurry into the fractured well.
  • 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.
  • Example 3 The location of the test site was selected to provide a compact shale of suflicient thickness and depth to give a useful test.
  • the injection well was drilled to approximately 1100 feet, cased with -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 2 /2" 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.
  • 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.
  • a method of treating radioactive waste solutions and slurries for disposal which comprises, adding to said waste clay minerals, lime and caustic, wherein the clay minerals are added in an amount varying from 1 to 60 percent based on the weight of the total mixture, and wherein said lime is added in an amount varying from 2 to 30 percent based on the weight of said waste fluids, and said caustic being added in an amount sufficient to provide a pH of the mixture above about 10 and allowing said treated waste to solidify.
  • a method of treating radioactive waste solutions and slurries for disposal which comprises, adding to said waste clay minerals, lime and caustic, wherein the clay minerals are added in an amount varying from 1 to 60 percent based on the weight of the total mixture, and wherein the amount of lime added is sufiicient to give a Ca ion concentration varying from /2 to 25 percent based on the total weight of all components exclusive of the caustic, and said caustic being added in an amount sufficient to provide a pH of the mixture varying from 10 to 13 and allowing said treated waste to solidify.
  • a method of treating radioacitve waste solutions and slurries for disposal which comprises, adding to said waste natural clay, lime and caustic, wherein said natural clay is added in an amount varying from 20 to 60 percent based on the weight of the total mixture, and wherein lime is added in an amount varying from 2 to 30 percent based on the weight of said waste fluids, and said caustic being added in an amount sufficient to provide a pH of the mixture varying from 10 to 13 and allowing said treated waste to solidify.
  • a method of treating radioactive waste solutions and slurries for disposal which comprises, adding to said waste natural clay, lime and caustic, wherein said natural clay is added in an amount varying from 25 to 45 percent based on the weight of the total mixture, and wherein lime is added in an amount from 3 to 15 percent based on the weight of said Waste fluids, and said caustic being added in an amount sufiicient to provide a pH of the mixture varying from 10 to 13 and allowing said treated waste to solidify.
  • a method of treating radioactive waste solutions and slurries for disposal which comprises, adding to said waste natural clay, lime and caustic, wherein said natural clay is added in an amount varying from 25 to 45 percent based on the weight of the total mixture, and wherein the amount of lime added is sufficient to give a Ca ion concentration varying from /2 to 25 percent based on the total weight of all components exclusive of the caustic, and said caustic being added in an amount sufficient to provide a pH of the mixture varying from 10 to 13 and allowing said treated waste to solidify.
  • a method of treating radioactive waste solutions and slurries for disposal which comprises, adding to said waste natural clay, lime and caustic, wherein said natural clay is added in an amount varying from 25 to 45 percent based on the weight of the total mixture, and wherein lime is added in an amount from 3 to 15 percent based on the weight of said waste fluids, and said caustic being added in an amount sulficient to provide a pH of the mixture varying from 10 to 13 and allowing said treated waste to solidify.
  • a method of disposal of radioactive waste solutions and slurries in fixed fashion and in solid form which comprises, adding clay minerals, lime and caustic to said radioactive waste, said clay minerals being added in an amount varying from 1 to 60 percent based on the weight of the total mixture, said lime being added in an amount varying from 2 to 30 percent based on the weight of said waste, and said caustic being added in an amount sufficient to provide a pH varying from 10.5 to 12, and then subjecting the final mixture to an in place temperature in the range of approximately 65 to 500 F. and allowing the treated waste to solidify in place.
  • a composition of matter which solidifies upon standing at temperatures of approximately 65 F. and above comprising radioactive waste selected from the group consisting of radioactive solutions and radioactive slurries, water, clay minerals, lime and caustic, said clay minerals being present in an amount varying from 1 to 60 percent based on the weight of the total mixture, said lime being present in an amount varying from 2 to 30 percent based on the weight of said waste, and said caustic being present in an amount sufficient to give the mixture a pH varying in the range of to 13.
  • a composition which solidifies upon standing at temperatures of approximately 65 F. and above comprising radioactive waste selected from the group consisting of radioactive solutions and radioactive slurries, water, clay minerals, lime and caustic, said clay minerals being present in an amount varying from 1 to 60 percent based on the weight of the total mixture, said lime being present in an amount sufficient to give a calcium ion concentration of /2 to 25 percent based on the total weight of all components exclusive of the caustic, said caustic being present in an amount sufiicient to provide a pH bearing from 10 to 13.
  • composition of claim 8 wherein the clay minerals are natural clay and wherein said clay is present in an amount varying from 20 to 60 percent.
  • composition of claim 8 wherein said clay mineral is natural clay and is present in an amount varying from 25 to 45 percent, and wherein said lime is present in an amount varying from 3 to percent.
  • composition of claim 9 wherein the clay mineral is natural clay and is present in an amount varying from 25 to 45 percent.
  • composition of claim 11 wherein the caustic is present in an amount sufiicient to provide a pH of the mixture varying from 10.5 to 12.
  • a method of subterranean disposal of radioactive waste which comprises providing a wellbore which penetrates an impermeable rock formation, fracturing the impermeable rock formatiton through the wellbore in a substantially horizontal direction to provide at least one fracture therein at a depth sufficient to provide shielding at the surface from the most energetic fraction of radioactive waste to be injected thereinto, said fracture being confined within said impermeable rock formation, injecting a solidifiable material containing radioactive waste into a fracture so formed in said impermeable rock formation and thereafter injecting additional solidifiable material free of radioactive waste to seal the radioactive waste in said impermeable rock formation.
  • a method of subterranean disposal of radioactive waste which comprises providing a wellbore which penetrates an impermeable rock formation, fracturing the impermeable rock formation through the wellbore in a substantially horizontal direction to provide at least one fracture therein at a depth sufficient to provide shielding at the surface from the most energetic fraction of radioactive waste to be injected thereinto, said fracture being confined within said impermeable rock formation, injecting a stream of solidifiable material free from radioactive waste into a fracture so formed in said impermeable rock formation; thereafter introducing radioactive waste to said stream, subsequently discontinuing the introduction of radioactive waste and continuing the injection of solidifiable material free from radioactive waste to seal the radioactive Waste in said impermeable rock formation.
  • the solidifiable material comprises clay minerals in an amount varying from 1 to percent based on the weight of the total mixture (including the radioactive waste), lime is in an amount varying from 2 to 30 percent based on the weight of the radioactive waste, and caustic in an amount sufficient to provide a pH of the mixture (including the radioactive waste) above about 10.
  • a method of subterranean disposal of radioactive waste which comprises providing a wellbore which penetrates an impermeable rock formation, fracturing the impermeable rock formation through the wellbore to provide at least one fracture therein at a depth sufiicient to provide shielding at the surface from the most energetic fraction of radioactive waste to be injected thereinto, said fracture being confined within said impermeable rock formation, injecting a solidifiable material containing radioactive waste into a fracture so formed in said impermeable rock formation and thereafter injecting additional solidifiable material free of radioactive waste to seal the radioactive waste in said impermeable rock formation.

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NL271326D NL271326A (en, 2012) 1958-12-31
GB36812/59A GB926822A (en) 1958-12-31 1959-11-13 Method of fixing atomic wastes for disposal
FR814023A FR1243673A (fr) 1958-12-31 1959-12-23 Procédé de fixation des résidus atomiques pour leur mise au rebut
BE597015A BE597015Q (fr) 1958-12-31 1960-11-14 Procédé de fixation des résidus atomiques pour leur mise au rebut
US88246A US3274784A (en) 1958-12-31 1961-02-24 Composition and method for fixing atomic waste and disposal
FR871221A FR1298142A (fr) 1958-12-31 1961-08-21 Procédé de mise au rebut des résidus radio-actifs et composition correspondante
GB40762/61A GB940254A (en) 1958-12-31 1961-11-14 Improvements in or relating to the disposal of radioactive waste

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US3482634A (en) * 1968-06-28 1969-12-09 Milchem Inc Process for sealing porous earth formations by cementing
US3935098A (en) * 1972-07-04 1976-01-27 Nippon Soda Co., Ltd. Adsorbent process for heavy metals
US4028265A (en) * 1974-04-02 1977-06-07 The United States Of America As Represented By The United States Energy Research And Development Administration Process for converting sodium nitrate-containing, caustic liquid radioactive wastes to solid insoluble products
US4054320A (en) * 1976-08-24 1977-10-18 United States Steel Corporation Method for the removal of radioactive waste during in-situ leaching of uranium
US4087375A (en) * 1975-05-07 1978-05-02 Shin Tohoku Chemical Industry Co., Ltd. Method for treating radioactive waste water
US4146095A (en) * 1977-07-15 1979-03-27 Alspaw D Ivan Method and apparatus for nuclear heating of oil-bearing formations
US4528129A (en) * 1982-05-03 1985-07-09 Frank Manchak Processing radioactive wastes and uranium mill tailings for safe ecologically-acceptable disposal
US4560503A (en) * 1982-08-30 1985-12-24 John D. Gassett Fluid waste disposal
US4576513A (en) * 1981-10-22 1986-03-18 Wintershall Ag Process for terminal storage of pumpable wastes
US4591455A (en) * 1982-11-24 1986-05-27 Pedro B. Macedo Purification of contaminated liquid
DE3545592C1 (de) * 1985-12-21 1987-06-25 Nukem Gmbh Verfahren zur Konditionierung von wasserloeslichen Sonderabfaellen
US4737316A (en) * 1982-11-24 1988-04-12 Pedro B. Macedo Purification of contaminated liquid
US4775494A (en) * 1985-06-10 1988-10-04 Rowsell Farrell D Hazardous and radioactive liquid waste disposal method
US4778627A (en) * 1986-01-13 1988-10-18 James William Ayres Disposable hazardous and radioactive liquid hydrocarbon waste composition and method
US4781860A (en) * 1986-01-13 1988-11-01 James W. Ayres Disposable hazardous and radioactive liquid aqueous waste composition and method
US4950426A (en) * 1989-03-31 1990-08-21 Westinghouse Electric Corp. Granular fill material for nuclear waste containing modules
US5075044A (en) * 1986-07-07 1991-12-24 Electricite De France Service International Process for the radioactive decontamination of an oil
US5098612A (en) * 1988-12-10 1992-03-24 Rowsell Farrell D Method of preparing solidified and stabilized hazardous or radioactive liquids
USRE33915E (en) * 1986-01-13 1992-05-05 James William Ayres Disposable hazardous and radioactive liquid hydrocarbon waste composition and method
USRE33955E (en) * 1985-06-10 1992-06-09 Hazardous and radioactive liquid waste disposal method
USRE34041E (en) * 1986-01-13 1992-08-25 James William Ayres Disposable hazardous and radioactive liquid aqueous waste composition and method
US5141365A (en) * 1988-07-14 1992-08-25 Fosroc International Limited Backfilling in mines
US5202522A (en) * 1991-06-07 1993-04-13 Conoco Inc. Deep well storage of radioactive material
US6284681B1 (en) 1999-03-05 2001-09-04 Westinghouse Savannah River Company Reactive composite compositions and mat barriers
US6973758B2 (en) 2001-05-14 2005-12-13 Rad Technology, Llc Shielded structure for radiation treatment equipment and method of assembly
US20080004477A1 (en) * 2006-07-03 2008-01-03 Brunsell Dennis A Method and device for evaporate/reverse osmosis concentrate and other liquid solidification
US7361801B1 (en) 2003-08-27 2008-04-22 352 East Irvin Avenue Limited Partnership Methods for immobilization of nitrate and nitrite in aqueous waste
WO2013003796A1 (en) * 2011-06-29 2013-01-03 GERMANOVICH, Leonid Abyssal sequestration of nuclear waste and other types of hazardous waste
EP2913825A4 (en) * 2012-10-25 2016-05-04 Vladimir Nikolaevich Ivanov METHOD FOR PRODUCING AND TEMPERING RADIOACTIVE WASTE
US9630225B2 (en) 2014-01-28 2017-04-25 Red Leaf Resources, Inc. Long term storage of waste using adsorption by high surface area materials
US20180075935A1 (en) * 2016-09-12 2018-03-15 Grand Abyss, Llc Emergency method and system for in-situ disposal and containment of nuclear material at nuclear power facility
US10807132B2 (en) 2019-02-26 2020-10-20 Henry B. Crichlow Nuclear waste disposal in deep geological human-made caverns

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FR1438454A (fr) * 1965-03-30 1966-05-13 Commissariat Energie Atomique Perfectionnements aux procédés pour injecter des déchets radio-actifs dans le sol
DE2917060C2 (de) * 1979-04-27 1983-10-27 Kernforschungszentrum Karlsruhe Gmbh, 7500 Karlsruhe Verfahren zur Verfestigung von tritiumhaltigem Wasser
FR2516292A1 (fr) * 1981-11-10 1983-05-13 Stockage Assainissement Coulis special d'injection et son utilisation pour le stockage dans le sol de dechets radioactifs
DE3219114A1 (de) * 1982-05-21 1983-11-24 Kernforschungsz Karlsruhe Verfahren zur verbesserung der eingenschaften von verfestigungen radioaktiver festabfaelle
DE3840794A1 (de) * 1988-12-03 1990-06-13 Battelle Institut E V Verfahren zur herstellung eines festen abfallproduktes zur endlagerung radioaktiver stoffe
RU2307412C2 (ru) * 2003-10-24 2007-09-27 Федеральное государственное унитарное предприятие "Сибирский химический комбинат" Министерства Российской Федерации по атомной энергии Способ захоронения жидких радиоактивных отходов
RU2504850C1 (ru) * 2012-05-15 2014-01-20 Валерий Иванович Сергеев Способ консервации приповерхностного хранилища, содержащего радиоактивные отходы и устройство для его реализации
CN104464867B (zh) * 2014-12-03 2017-03-15 中国工程物理研究院材料研究所 一种放射性废机油高强度水泥固化体的制备方法
RU2632801C1 (ru) * 2016-11-03 2017-10-09 Елена Васильевна Захарова Способ глубинного захоронения облученного графита уран-графитовых ядерных реакторов
CN109524144B (zh) * 2018-12-11 2022-04-08 湖南理工学院 一种低放废油的固化处理配方

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Cited By (49)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
US3379013A (en) * 1964-12-29 1968-04-23 Halliburton Co Method of disposing of waste materials
US3482634A (en) * 1968-06-28 1969-12-09 Milchem Inc Process for sealing porous earth formations by cementing
US3935098A (en) * 1972-07-04 1976-01-27 Nippon Soda Co., Ltd. Adsorbent process for heavy metals
US4028265A (en) * 1974-04-02 1977-06-07 The United States Of America As Represented By The United States Energy Research And Development Administration Process for converting sodium nitrate-containing, caustic liquid radioactive wastes to solid insoluble products
US4087375A (en) * 1975-05-07 1978-05-02 Shin Tohoku Chemical Industry Co., Ltd. Method for treating radioactive waste water
US4054320A (en) * 1976-08-24 1977-10-18 United States Steel Corporation Method for the removal of radioactive waste during in-situ leaching of uranium
US4146095A (en) * 1977-07-15 1979-03-27 Alspaw D Ivan Method and apparatus for nuclear heating of oil-bearing formations
US4576513A (en) * 1981-10-22 1986-03-18 Wintershall Ag Process for terminal storage of pumpable wastes
US4528129A (en) * 1982-05-03 1985-07-09 Frank Manchak Processing radioactive wastes and uranium mill tailings for safe ecologically-acceptable disposal
US4560503A (en) * 1982-08-30 1985-12-24 John D. Gassett Fluid waste disposal
US4737316A (en) * 1982-11-24 1988-04-12 Pedro B. Macedo Purification of contaminated liquid
US4591455A (en) * 1982-11-24 1986-05-27 Pedro B. Macedo Purification of contaminated liquid
US4775494A (en) * 1985-06-10 1988-10-04 Rowsell Farrell D Hazardous and radioactive liquid waste disposal method
USRE33955E (en) * 1985-06-10 1992-06-09 Hazardous and radioactive liquid waste disposal method
DE3545592C1 (de) * 1985-12-21 1987-06-25 Nukem Gmbh Verfahren zur Konditionierung von wasserloeslichen Sonderabfaellen
USRE33915E (en) * 1986-01-13 1992-05-05 James William Ayres Disposable hazardous and radioactive liquid hydrocarbon waste composition and method
US4781860A (en) * 1986-01-13 1988-11-01 James W. Ayres Disposable hazardous and radioactive liquid aqueous waste composition and method
US4778627A (en) * 1986-01-13 1988-10-18 James William Ayres Disposable hazardous and radioactive liquid hydrocarbon waste composition and method
USRE34041E (en) * 1986-01-13 1992-08-25 James William Ayres Disposable hazardous and radioactive liquid aqueous waste composition and method
US5075044A (en) * 1986-07-07 1991-12-24 Electricite De France Service International Process for the radioactive decontamination of an oil
US5141365A (en) * 1988-07-14 1992-08-25 Fosroc International Limited Backfilling in mines
US5098612A (en) * 1988-12-10 1992-03-24 Rowsell Farrell D Method of preparing solidified and stabilized hazardous or radioactive liquids
US4950426A (en) * 1989-03-31 1990-08-21 Westinghouse Electric Corp. Granular fill material for nuclear waste containing modules
EP0390375A2 (en) 1989-03-31 1990-10-03 Westinghouse Electric Corporation Granular fill material for nuclear waste containing modules
US5202522A (en) * 1991-06-07 1993-04-13 Conoco Inc. Deep well storage of radioactive material
US6284681B1 (en) 1999-03-05 2001-09-04 Westinghouse Savannah River Company Reactive composite compositions and mat barriers
US6973758B2 (en) 2001-05-14 2005-12-13 Rad Technology, Llc Shielded structure for radiation treatment equipment and method of assembly
US7361801B1 (en) 2003-08-27 2008-04-22 352 East Irvin Avenue Limited Partnership Methods for immobilization of nitrate and nitrite in aqueous waste
US20080004477A1 (en) * 2006-07-03 2008-01-03 Brunsell Dennis A Method and device for evaporate/reverse osmosis concentrate and other liquid solidification
CN106782735B (zh) * 2011-06-29 2018-08-31 宏大深渊有限责任公司 核废料和其他类型的有害废料的深埋
US10032535B2 (en) 2011-06-29 2018-07-24 Grand Abyss Llc Abyssal sequestration of nuclear waste and other types of hazardous waste
CN103814411A (zh) * 2011-06-29 2014-05-21 宏大导航有限责任公司 核废料和其他类型的有害废料的深埋
JP2014522975A (ja) * 2011-06-29 2014-09-08 グランド・ダイレクションズ・エルエルシー 核廃棄物および他の有害廃棄物の深層隔離
US9190181B2 (en) 2011-06-29 2015-11-17 Grand Directions, Llc Abyssal sequestration of nuclear waste and other types of hazardous waste
EP3660866A1 (en) * 2011-06-29 2020-06-03 Grand Abyss LLC Abyssal sequestration of nuclear waste and other types of hazardous waste
CN103814411B (zh) * 2011-06-29 2016-12-14 宏大深渊有限责任公司 核废料和其他类型的有害废料的深埋方法及系统
JP2017062245A (ja) * 2011-06-29 2017-03-30 グランド・アビス・エルエルシー 核廃棄物および他の有害廃棄物の深層隔離
US10629315B2 (en) 2011-06-29 2020-04-21 Grand Abyss, Llc Abyssal sequestration of nuclear waste and other types of hazardous waste
CN106782735A (zh) * 2011-06-29 2017-05-31 宏大深渊有限责任公司 核废料和其他类型的有害废料的深埋
US9700922B2 (en) 2011-06-29 2017-07-11 Grand Abyss, Llc Abyssal sequestration of nuclear waste and other types of hazardous waste
US9741460B2 (en) 2011-06-29 2017-08-22 Grand Abyss, Llc Abyssal sequestration of nuclear waste and other types of hazardous waste
WO2013003796A1 (en) * 2011-06-29 2013-01-03 GERMANOVICH, Leonid Abyssal sequestration of nuclear waste and other types of hazardous waste
KR20140056243A (ko) * 2011-06-29 2014-05-09 그랜드 다이렉션스, 엘엘씨 핵 폐기물 및 다른 유형의 유해 폐기물의 심해 격리
EP2913825A4 (en) * 2012-10-25 2016-05-04 Vladimir Nikolaevich Ivanov METHOD FOR PRODUCING AND TEMPERING RADIOACTIVE WASTE
US9630225B2 (en) 2014-01-28 2017-04-25 Red Leaf Resources, Inc. Long term storage of waste using adsorption by high surface area materials
US20180075935A1 (en) * 2016-09-12 2018-03-15 Grand Abyss, Llc Emergency method and system for in-situ disposal and containment of nuclear material at nuclear power facility
US10115489B2 (en) * 2016-09-12 2018-10-30 Grand Abyss, Llc Emergency method and system for in-situ disposal and containment of nuclear material at nuclear power facility
US11270805B2 (en) 2016-09-12 2022-03-08 Grand Abyss, Llc Emergency method and system for in-situ disposal and containment of nuclear material at nuclear power facility
US10807132B2 (en) 2019-02-26 2020-10-20 Henry B. Crichlow Nuclear waste disposal in deep geological human-made caverns

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FR1243673A (fr) 1960-10-14
BE597015Q (fr) 1961-03-01
FR1298142A (fr) 1962-07-06
GB940254A (en) 1963-10-30
GB926822A (en) 1963-05-22
NL271326A (en, 2012)

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