CA1057947A - Process for solidifying nuclear materials - Google Patents

Abstract

ABSTRACT OF DISCLOSURE
Radioactive liquid materials or wastes are contacted with complex anions having the general formula (MxOyHz)-where M may be titanium, niobium, tantalum or zirconium, and the radionuclide ions in the material react and form a precipitate with the complex anions. This precipitate, upon separation and subsequent beating to from about 100°
to about 1000°C, forms chemically stable, complex oxide mixtures.

Description

i~5 ~

8ACKGROUND OF ItlVENTION
The invention relate~ to a novel method for converting radioactive materials or waste in li~uid solution~ to radloactive materials or waste i~ a solid, chemically stable form.
Reactor operation and reactor fuel reprocessing operations result in radioactive materials or waste which generally is in the form of aqueous solutions. Conversion of this material to a solid form is required in order to minimize transportation and ~torage problems since solidification reduce~ the mobility of the material, re~ults in a volume reduction of about ten-fold, and provides greater protection in terms of release of material to the surroundings. It is desirable that the solid material or waste be in a chemically stable or inert form to substan-tially increase the safety as~ociated wlth the storing and disposal of these materials.
Efforts to com~ert radioactive waste from an aqueous solution to a solid form have resulted in a variety of proces~es such as d~scribed by K. J. Schneider in Reactor TechnologY, "Solidification and Disposal of High-Level Radioactive Wastes in the United St8te8~, Vol. 13, No. 4, Winter 1970-1971. These processes gcnerally include an evaporation step to remove water and a calcining step to volatilize some waste constituents, for e~ample, nitrate and sulfate ions. The material resulting from these processes may be an oxide mixture, or may be further incorporated in a glass matrix to reduce the high leaching rates exhibited by the oxide mixtures when in contact with water. The stability of the glass containing the radio-active waste ha~ been shown to be dependent upon the composition of the radioactive waste involved, and may differ from one bstch to another as well as from one repr~cessing site to another. Further, other disadvantages of these solidification processes ~nd the solid waste produced include low thermal conductivity of the product, compllcated system requirements, necessity for good flow control of difficult to handle feed solution~, etc, SUMM~RY OF INVENTION
In view of the above, it is an ob~ect of this invention to provide a novel process for solidifying materials such as radioactive liquid waste which employs complex anions having the general formula (MxOyHz) to react with cations in these materials and form precipitates therewith, It is further ob~ect of this invention to provide an operationally simple process for converting radioactive waste in the form of dissolved salts ln aqueous solutions to radioactive waste in a solid form.
It is a further ob~ect of thls invention to provide a process for removing radioactive waste from aqueous solu-tions without an evaporation step to remove the water.
It is a further ob~ect of this invention to provide a - - ', .

i~5 ~
process for converting radioacti~e waste in aqueous solutions to a chemically stable form that i~ solid, which process does not require a separate calcining step to volatilize waste constituents such 88 the nitrate and sulfate ions.
It is a further ob~ect of this invention to provlde a process for conYerting radioactive waste in aqueous solutions to a chemically stable ~olid form without carrying into the ~olid form constituents such as nitrate ions and sulfate ions.
It is a further ob~ect of this invention to provide a simple method for converting radioactive waste in aqueous solutions into a chemically stable solid form wherein separation factors of equal to or greater than about 10 may be achieved for the radioactive materials.
Various other ob~ects and advantages will appear from the following description of this inveation and the most novel features will be particularly pointed out hereinafter in connection with the appended claims. It will be under-stood that various changes in the dotail~, materials and steps of the proce~ses which are herein described and illu~trated in order to explain the nature of the invention may be effected by those skilled in the art without departing from thc scope of this invention.
The invention comprises contacting a ched cal compound containing a complex anionic group which has the 1 ~5~

general formula (~ O~z) , where M i8 titanium, niobium, tantalum or zirconlum and x i8 at least two, O i9 oxygen and H i~ hydrogen, and cations such ~ quaternary ammonium, ammonium, potassium, magnesium, h~drogen, sodium, lithium, etc., with aqueous radioactivc waste materials from reactor operations, reactor fuel reprocessing operations, and the like, to form water insoluble materials containing the radioactive co~ponents. These materials may be subsequently h-ated at from about 100C to about 1000C to form chemlcally stable complex oxide mixtures containing the radiosctive material ions.

DESCRIPTION OF DRAWING
.
Fig. 1 illustrate~ a processing sequence for using this invention.
Fig. 2 illustrates an alternate processing sequence.
, DETAILRD DESCRIPTION
As shown in Fig. 1, an aqueous waste solution containing radionuclides or radioactive elements, as shown in Table I, which may be gonerated from reactor operatiQns, reactor fuel reprocessing operations, or the like, is contacted with a chomical compound of general formula M'(MxOyHz) This invention is equally applicable to low level, int~rmediate level or high level radioactive materials or waste solutions, but is especially important for high level radioactive waste treatment. This compound may be produced .' : .

as described herein'oel~w. M7 repre~ent~ a cation ~uch as quaternary ammonium (QA+), hydrogen (H+), sodium (Na+), lithium (Ll ), ammonium (NH4 ), ~gnesium (Mg ), potassium (K+3, aluminum (Al) and other ions. In the anionic group or species of general formula (MkOyHz) , M is a metal such as titanium, niobium, tantalum, or zirconium, and x i8 at least two. The valence of the nionic specie~ i~ believed to be -1 from the manner in which the material reacts, i.e., ~toichiometr-Lc considera-tions, but the invention is not to be restricted to thi~
limitation.
Substituting into the general formula above gives four anionic species or classes (MkOyHz) which are (Ti205H) , (Zr2sH) , (Nb26H) and (Ta206H) .
For convenience of description, these are herein referred $o as titanate, zirconate, niobate, and tantalate ions respectively. The exact structural configuration of the empirical formula (MkOyHz) i~ not positively known and, as such, the correct chemical nomenclature is not provided. This in no way acts as an obstacle to an accurate description of this invention. The selection of the material Ml(MXOyHz) to be used is made from a consideration of cost, reactivity, interference with reaction, ease of removal from the precipitate that forms, and the like.

~ABLE I
Ma~or Fission Products in Radioactive Wastes Selenium Palladium Cerium Rubidium Sil~er Praesodymium Strontium Cadmium Neodymium Yttrium Indium Promethium Zirconium Tin Samarium Niobium Antimony Europium Molybdenum Tellurium Gadolinium Technetium cesium Terbium ~uthenium Barium Dysprosium Rhodium Lanthanum Compounds containing these anionic species may be produced through various reactions, uch as, in one example, the reaction of a quatern ry ammonium base R4NOH (where R
i8 any of several organic groups such as alkanes having from 1 to about 7 ~arbon atoms, aryl groups, etc. as known in the art) with a metal alkoxide (R'O)nM where n i8 e~ual to 4 when M is titanium or zirconium, or n is equal : 20 to 5 when M i8 niobium or tantalum, and R' is an alkyl group having from 1 to about 6 carbon atom~. Examples of the quaternary ammonium base may be such as tetramethyl ammonium hydroxide, (CH3)4NOH, tetrapentyl ammonium hydroxide [(CsHll)4NOH], and the like, and examples of the metal alkoxide may be such as tetraisopropyl titanate lV~

[(C3H70)4Til or pentaethyl niobate ~(C2H50)~Nb].
It is believed ~quation 1 ls a repre~entativeexpression for one reactlon for produclng the anion~c species of interest. It is here lntended to provide only a representative equation and not a balanced equation.

(CH3)4NOH + (C3H70)4Ti _ (CH3)4N(T120sH) + C3H70H (1) The (CH3)4N(Ti205H) product is referred to, for con-venience as described above, as tetramethyl ammonium titanate. The base may be added in an alcohol solution such as 25Z by weight of the base in methanol. As may be readily seen, the (OH) group appears to be replaced in the base with an anionic group of empirical formula (MxOyHz) 1. As such, the quaternary group may be alkanes, aryl groups, or many others since the~e do not enter directly into the reaction.
The quaternary ammonium titanate product produced in equation 1 may be used for the nuclear waste solidification reaction, or, if desired, the titanate prodùct may be contacted w~th a suitable salt or acid to replace the quaternary ammonium group with the cation of the salt or acld. It is believed that equation 2 is representative of this reaction.

R4N(~ OyHz) + NaN03 ~ R4N N03 + ~2) Na(~ OyHz) (precipitate) 1~3~ t~

Other salt cations that may be used in equation 2 are such as ammonium, lithlum, magnesium, aluminum, potassium, etc., together wi~h salt anionic groups such as chlorides, nitrates and sulfates. Some of the acids that may be used are such as nitric, hydrochloric, sulfuric, etc. The precipitate product of equation 2 can be readily separated by such as filtration or the like~
Another process for arriving at the (MxOyHz)~l group without the u8e of quaternary ammonium is believed to be represented by equation 3.

NaOH ~ (R'O) M Methanol n Nonaqueous Na(MxOyHz) (precipitate) + other products The process of equation 3 is more economical in arriving at the (MXOyHz) 1 anionic species than that described by equation 2 since the use of a quaternary ammonium material is avoided. Other bases that may be e~ployed are such as lithium hydroxide, potassium hydroxide, magnesium hydroxide, etc. The reaction is conducted in a nonaqueous alcohol bath.
In the process of contacting a base with a metal alkoxide, such as illustrated by equation 1 and 3. The mole ratio of base to alkoxide should be equal to or less than about 1/2. As thi~ ratio is éxceeded and approaches one, for example, the reaction products do not perform satisfactorily. This mole ratio of base to alkoxide may 1 ~5 7~

be substantially smaller, such as about 1/20 or le~s and the de~ired reaction products shown in equations 1 and 3 are still attained. t Sodium titanate (NaTi205H~ wa~ prepared using first the processes illustrated by equations 1 and 2 and secondly by the process of equation 3. In the first process, 284,6 grams of tetraisopropyl titanate were added to 182.0 grsms of a solution that is 25 percent by weight of tetramethylammonium hydroxide in methanol in a beaker.
SOO ~lliliters (ml) of deionized water were added and the mixture boiled until foaming began whereupon the mixture was then cooled to room temperature. One liter of 2 molar sodium nitrate ~olution was then added with stirring and, after thirty minutes, the supernatant solution was decanted. A liter of 0.5 molar sodium nitrate solution was then added, stirred, and after 10 minutes again decanted.
The residue was filtered and washed with two 200 ml.
~ volumes of deionized water and two 200 ml. volumes of ; acetone. The sodium titanate product wa~ then dried at 105C to constant weight.
In the second process, 20.0 gram~ of sodium hydroxide were dissolved in 2go ml. of methanol. The solution was filtered and 284.6 grams of tetraisopropyl titanate were added to the sodium hydroxide. One liter of deionized water was added to the reaction product sodium titanate, the mixture stirred, and the supernatant decanted. The _ g _ .
" : ' 1~3~ '7 residue was filtered and washed with two 200 ml. volumes of acetone and subsequently dried, The second process described lmmediately hereinabove may be modified slightly to yield a material that more readily lends itself to ion exchange column use because it has a much greater surface area and swelling and plugging problems which might have previously been encountered are significantly reduced. The modlfication comprises, in lieu of the one liter addition of deloniæed water to the reaction product, the addition of 1 1/2 liters of acetone containing 100 ml. of water. The other process steps remain the same. The sodium titanate product produced by the process employing this modification contains a much greater surface area than previously obtained, and permits the ion exchange process that removes radionuclides from solution using ion exchange column beds exclusively without separate batch equilibration.
Many compounds have been successfully used in the solidification process of nuclear waste material. A few of these compounds include NaTi205H, NH4Ti205H, (CH3)4NTi25H. NaZr205H, NaNb26H, HTi25H, Al(Ti25H)3, Li(Ti2osH)~ etc-Contacting of the chemical compound (for sake ofconvenience hereinafter we use sodium titanate, NaTi205H, as the representative chemical compound unless otherwise specified) with the aqueous solution containing the , ~,f~j'f'~'7 radioactive elements in batch eqwilibration (i.e., disposing an excess amount such as about 10% of ~aTi205H
with a known amount of radioactive waste solution in a container or tank with stirring until reaction is complete) results in a chemical reaction in which the complex anionic species or group reacts with the radio-active metal cations as well as some radionuclide anions to form water insoluble materials or precipitates including the radioactive materials. Contacting should be for at least about 15 minutes. The temperature of the waste material used has been about ambient temperature (about 26C) for this reaction.. Thc precipitate or solid formation is separated from the liquid by a suitable process known in the art such as filtration, centrifugation, and the like.
The use of the group (MxOyHz) in these exchange processes also effects removal of anionic radionuclides known to exist in anionic form such as ruthenlum, molybdenum, and tellurium. It is not understood why anions also exchange to form the precipitate, with the (MxOyHz) group, but this exchange and precip~tation does occur.
It ls believed that equation 4 represents a reaction that occurs to form the insoluble precipitate product.
Here the radionucllde may not only be a positive uni- or multivalent cation, but as described above, may also be a uni- or a multivalent anion.

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(radionuclide3 + (~ OyHz) (Radionuclide) ~ OyHz) (precipitate) The precipitate product of equation 4 is a solid inert material which contains the radioactive elements.
Since it is, as formed, in granular or powdery form, it may be desirable to treat the precipitate to remove moisture content and stabilize its leaching characteristics as well as to form it into a shape that is more easily handled and stored. To accomplish this, the precipitate may be heated to remove water content, heating may preferably be from about 100C to 400 or 500C and heating may be accomplished at ambient pressure and environment, or under a suitable reduced pressure, or in an inert or reducing atmosphere such as argon gas, nitrogen gas or carbon monoxide gas.
Heating to the temperatures recited permits the water to be removed and yet does not sinter the product.
Length of heating time will be dependent upon the heating temperature, whether in a reduced pressure environment or not, the quantity of material to be dried, etc. It may be desired to heat for not less than one hour. Heating to le~s than sintering temperature in an inert or reducing atmosphere has the further advantage that metals susceptible to oxidation may not be oxidized. Oxidation of these metals may adversely affect leaching characteristics after sintering, 1 ~5~

After the wa~er i8 removed, the dried powder or granules may be comminuted, if des~red, or u~ed as i~.
The precipitate i~ placed in a di~ or mold and compressed at a suitable pressure such as from about 3,000 to about 20,000 pounds per square inch (psi). The pressed and shaped form ls subsequently placed in a mold and heated to from about 600C to about 1000C, and held at temperature for from about 1 to a~out 2 hours to sin~er the ceramic material into a form that is safe to transport and store.
Selection of the exchange cation to be used in the chemical compound may enable the fractional separation or selective removal of particular radionuclides. For example, it may be shown that in~oluble lon exchange compounds formed with a trivalent cation may be used to ~egregate fission products and actinides based on valence differences. We believe equation 5 illustrates this fractional separation:

Cs (aqueous) + 3U02+2 (aqueous) + 2Al(Ti205H)3 (solid) _ -Cs+ (aqueous) + 2Al+3 (aqueous) + 3 U02(Ti205H)2 (solid) Since cesium and strontium radionuclides initially create the greatest hazard in the waste solution, and after about 300 years present no or minimal hazard, it may be desirable to first extract the other radionuclides from the waste solution, store the remaining cesium and strontium containing ~olution in an area where they can beappropriately maintained, and store the precipitated remaining radionuclides using the teaching~ of this invention as an inert product in underground storage or the like.
The reactions described herein with the multivalent metal cations are essentially quantitative precipitation reactions. Reactions with univalent cations are efective but may not result in a quantitative precipita-tion in one batch equillbratlon contacting step, but may do so with subsequent contacting ~s will be described hereinbelow. If, after batch equilibration and precipitate separation the solution from which the precipitate is removed is processed through a column ion exchange which has a Na(Ti205H) material or the like e~change bed, a measured separation factor for ce~ium-137 and plutonium-238 of greater than or equal to 10 results. The dotted line sequence in Fig. 1 illustrates a processing sequence that may be employed to accomplish this removal.
The cations of the chemical compounds which may be used do not enter directly into the chemistry of the solidification reaction, but do affect the process as relates to the intermediate forms achieved as described hereinbelow.
Quaternary ammonium compounds or other compounds as described above which have the de~ired anionic species i~5~
(MkOyHz) and which are water soluble may be added to radioactive waste in an aqueous waste solution as an aqueous liquid or solution. Reaction of the anionic species with the radioactive elements and other cationic elements occurs immediately and results in the formation of a gelatinou~ precipitate, it is believed as noted in equation 6, which may be separated from the liquid by freezing the mixture and, after su~sequent thawing, filtering or otherwise processlng the mix~ure to effect separation of the precipitate.

Cs (aqueous) + (Ti205H) (aqueous) ~ (6) Cs~Ti205H) (gelatinous precipitate) The compound having the de~ired anionlc species may also be in the solid form such as a quaternary ammonium substituted titanate such as NH4Ti205H or LiTi205H, or may be a solid quaternary ammonium compound, and may be added directly to the waste solution to effect the precipitation reaction, it is believed as noted in equation 7, although at slower rate, but the resulting precipitate may be easily filtered from the remaining, unreacted solution. Equation (7) employs cesium and tetramethyl ammonium titanate ((CH3)4NTi205H):

cs+ (aqueous) + ~CH3)4NTi20sH(Slid) r CsTi205H (Solid) + (CH3)4N (aqueous) S ~JL'7~ ;J

In this latter equation, a particular application for thls invention ls suggested -i.e.- tne removal of fission products from high concentration sodium salt solutions found in present tank stored wastes. It is beliæved that equation (8) represents this reaction.

Cs+(aqueous) + Na (aqueous) + NaNb206H(Solid) _ (8) Na (aqueous) + CsNb206H(Solid) It is theorized that the equations illustrated herein describe the reactions taking place but applicant is not to be restricted to this theory.
Water insoluble compounds formed by the process of equations 1, 2 and 3, may also be contacted with the aqueous waste solution containing the radioactive elements to effect an exchange reaction between the radionuclides in the aqueous solution and precipitate product of the equations. The new insoluble products may thereafter be separated from the solutions such as by filtration or the like.
Contacting as used herein describes the passing and mixing of the waste solution into a solution containing the (MxOyHz) groups, such as in batch equilibration, or may be used to describe such as passing a waste solution through a columnar bed of the NaTi205H or other like materisl through gravity flow or the like. Column ion exchange ~s possible by using insoluble materials containing titanates, ~ '7niobates, tantalates, or zirconates.
In some runs using this i~vention, the waste solution used contained 21 elements as listed in Table 2 which closely approximates the composition of a waste solution resulting from the reprocessing of power reactor fuel.
The procedures involved adding 250 ml. of the Table 2 ~olution to about 114 grams of NaTi205H, (which includes a 10% excess over the calculated stoichiometric amount required) and, in another run, adding to 250 ml. of the Table 2 solution about 87.4 grams of NH4Ti205H (which is again a 10% exces~ over the calculated stoichiometric amount required). In each case the solid and liquid were equilibrated for 1 hour while stirring before the solid was remoYed by filtering, washed with three 50 ml. volumes of water and dried for about 15 hours at about 105C.
Analysis of the waste solution after the solidification proce4s u6ing in one case NaTi205H ion exchange material and in another case NH4Ti205H, indicated that all multi-valent metal cations had been quantitatively removed (i.e., greater than 99.9%) and that significant reduction had occurred in concentrations of the univalent ions of rubidium and cesium, the reduction being about 50% in the case of rubldium and about 90X in the case of cesium.
A sample of each solid material removed from the above process was pressed into a cylindrical shape and calcined at about 1000C. X-ray diffraction studies of the fired ~5Y~

at 1000C reaction product of the NH4Ti205H-waste reaction revealed tltanium dioxide a~ the ma3Or phase present while the fired at 1000C reaction pro~uct of the NaTi205H-waste reaction revealed Na2Ti5011 as the ma~or phase present.
Leachlng, as used herein, refer6 to placing the material being leached under at least three inches of water, which water may be either static or dynamic. Leaching re3ults for waste reacted ion exchange material are given in Table 3 (daily test? and Table 4 (weekly test). The data obtained herein are comparable to minimum leach rates reported for the processes involving incorporation of waste in glass matrices. Table 3 indicates that after initial high leach rates, the leach rate stabilizes at a low level as shown in Table 4.

lV~ 34~' TABLE 2~

Number _lement Conc. (Molar) hydrogen ~ 1

2 iron 0.ûS

3 chromium 0.012

4 nickel 0.005 ~trontium 0.027 6 barium 0.027 7 cesium 0.054 ; 8 rubidium 0.01 9 yttrlum 0.01 lanthanum 0.01 .
11 cerium 0.028 12 praesodymium 0.013 13 europium 0.004 14 ruthenium 0.0099 rhodium 0.0097 16 palladium 0.0094 17 cadmium 0.002 18 uranium 0.011 19 molybdenum 0.025 tellurium 0.012 21 zirconium 0.106 *pH - 1.93 .c v ~ o ;~
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Comparison runs have also been conducted between the solidified product of this novel process as produced and also as mixed, consolidated, or b~ended with a filler material such as glass. To illustrate, samples were prepared from the powder form product of the (CH3)4NTi205H
and waste reaction whlch had been fired at 700C for about two hours. Thereafter, a first part of this product was cold pressed at about 4000 p8i in a one inch diameter die followed by firing for 5 hours at 900C in an air atmosphere and ambient pressure. A second part of the product was first mixed with 30 weight percent borosilicate glass and thereafter cold pressed and fired as done on the first part. Density and porosity data are prcvided in Table 5.

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m From the data of Table 5, it may be seen that although 30 weight percent (44 volume percent) glass was added to the waste, the volume required to contain a given quantity of waste remains substantially identical to the (CH3)4NTi2O5H - waste reaction product volume for the same quantity of waste. The glass addition decreases the overall porosity as well as markedly decreases the amount of open porosity, i.e., porosity wherein the pores are inter-connected. Consequently the glass bonded (CH3)4NTi205H -" ' .

... .

l~S~,s,~4~;, waste reactlon product exhibits a signlficantly higher thermal conductivity and a lower leach rate than the product without the glass.
These results illustrate that binder additions such as metals, glass or other ceramics may be advantageously u~ed in con~olidating or reducing porosity of the final solid wa3te product. High thermal conductivity is important in order to remove heat from the core or central portion of the stored wastes. Low lcach rates are lmportant in order to minimize the amount of radioactive material that is undesirably removed from the solidified waste.
Separation factors greater than 109 have been observed in radiochemical studies of 137-cesium and 238-plutonium and of other cation~. Most constituents of the aqueous waste solution containing radioactive elements are reduced below the limits of detection by traditional chemical techniques in a single batch equilibration process, and the separation factors of about 109 for fission products have been achieved using inorganic ion exchange columns employing the chemistry described hereinabove. The resulting titanate and waste reaction product has very low leach rates of about 10 6 to about 10 grams per square centimeter per day (g/cm -day).
Table 6 illustrates leach rate results of a (CH3)4NTi205H - waste reaction product which was fired at , -' ' : .
,: :

~ 1~5~
600C for 2 hours, pressed into a pellet at 4000 psi, and fired agaln at 1000C for 4 hour~. The results illustrate the lcw leach rates for the product obtained using the process of this inve~tion.
In addition to batch equilibration proce~ses, packed column ion exchange systems may also be employed as described aboYe and shown by the dotted line sequence of Fig. 1. Because of thls, a number of possible process configurations may be used for the removal of radioactive materials from radioactive liquid waste solutions. As has been previously di~cussed, a single stage of contacting or batch equilibration between the waste solution and the ion exchange materials removes about 99% of the multivalent radionuclides from the aqueous phase. Further contacting may be desired to remove the remaining 1% and a packed column process may be used.. The packed column ion exchange bed particles may preferably be from about 60 to about ; 200 mesh, although the use of sizes outside this range is still within the teaching of this invention. The combination of batch and packed column ion exchange allows separation from the waste material of a large portion of the radioactive cations followed by a "clean-up" column process that removes the remaining radloactive material from the waste solutions.

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An alternative proces~ may feed the waste solutlon directly into an ion exchange column without the need for batch equilibration as herein described. This contacting results in a separation factor of equal to or greater than for mNlti- and univalent cations.
Conventional regenerable ion exchange material such as cationic inorganic resins such as phenolic resins and polystyrene resins may be u~ed in the solidification ~ -proccss as diagrammatically illustrated by Fig. 2. This has the advantage of not having to empty the ion exchange column to remove the radionuclide (MxOyHz) material which would otherwise comprise the ion exchange column.
Regenerating solution passing through the organic resin ion exchange beds is fed to the waste stream as illustrated and the radionuclides removed from the packed columns thereby are brought into batch e~uilibration contact for removal of solidified radioactive waste. This waste may be further processed as described hereinsbove by heating, pressing, sintering, etc.
A particular advantage of the process described herein over some pr~or art processes is that calcining may be effected at between 600C and 1000C instead of at higher temperatures such as 1200-C, which higher temperature had the attendant problems of control of the radioactivity.

Claims (11)

What we claim is:

1. A process for removing radionuclide ions from radioactive aqueous solution into a solid chemically stable form comprising:
(a) reacting a metal alkoxide of general formula (R'O)nM with a base in a nonaqueous alcohol solution wherein M is niobium, titanium, tantalum, or zirconium, R' is an alkyl group having from 1 to about 6 carbon atoms, n is equal to the valence of M, said base is lithium hydroxide, sodium hydroxide, magnesium hydroxide, or potassium hydroxide, and the mole ratio of said base and said alkoxide is not greater than about 1/2;
(b) hydrolyzing the reaction product of step (a) to form an insoluble precipitate; and (c) contacting said radioactive aqueous solution with said insoluble precipitate to effect a reaction and yield a liquid solution and a chemically stable solid containing said radionuclide ions, and thereafter separating said stable solid from said liquid solution.

2. The process of claim 1 wherein said radioactive aqueous solution is from reactor fuel reprocessing operations and said radionuclide ions are of selenium, rubidium, strontium, yttrium, zirconium, niobium, molybdenum, technetium, ruthenium, rhodium, palladium, silver, cadmium, indium, tin, antimony, tellurium, cesium, barium, lanthanum, cerium, praesodymium, neodymium, promethium, samarium, duropium, gadolinium, terbium, or dysprosium.

3. The process of claim 1 wherein said insoluble precipitate is NaTi2O5H, NaZr2O6H, NaNb2O6H, or LiTi2O5H.

4. The process of claim 1 further including heating said stable solid at a temperature from about 100°C to about 500°C in an atmosphere of air, inert gas, or reducing gas at a reduced pressure to evaporate moisture retained in said stable solid.

5. The process of claim 4 including after said drying, compressing said stable solid to a desired shape at a pressure from about 3000 pounds per square inch to about 20,000 pounds per square inch, and thereafter heating said compressed shape at from about 600°C to about 1000°C for from about 1 to about 2 hours to form a chemically stable, inert, oxide ceramic product.

6. The process of claim 1 wherein said separating comprises filtering said stable solid, washing said filtered solid with a wash solution and again filtering said solid from said wash solution, and thereafter drying said solid.

7. The process of claim 1 wherein said contacting step comprises disposing said insoluble precipitate in a container with said radioactive aqueous waste solution and stirring said insoluble precipitate and said solution in said container.

8. The process of claim 6 including after said separating, passing said filtrate through an ion exchange column containing said insoluble precipitate as a water insoluble ion exchange bed material to remove radionuclide ions from said filtrate, separating said bed material from said filtrate, and subsequently heating said bed material to yield an inert oxide, product containing said radioactive elements.

9. The process of claim 8 wherein said ion exchange column together with said contacting provides separation factor for said radionuclide ions of greater than or equal to 109.

10. The process of claim 1 wherein said contacting comprises passing said radioactive aqueous solution through an ion exchange column containing said insoluble precipitate as a water insoluble ion exchange bed material.

11. The process of claim 1 wherein greater than about 99.9% of multivalent radioactive elements are removed from said aqueous solution into said chemically stable solid.