CA1091033A - Treatment of actinide-containing organic waste - Google Patents

Abstract

ABSTRACT
An actinide-containing organic waste is treated to achieve a substantial reduction in the volume of the organic waste and provide for recovery of the actinide constituent therefrom. The organic actinide-containing waste is reacted with a source of gaseous oxygen in a molten salt bath main-tained at an elevated temperature to produce gaseous reaction products comprising carbon monoxide and water vapor. The actinide and inorganic ash constituents of the waste are retained in the molten salt. Intermittently or continuously at least a portion of the molten salt is withdrawn and mixed with an aqueous medium to dissolve the salt constituents.
The medium then is filtered to remove the insoluble inorganic ash constituents and the actinide. The filter cake then is leached with an inorganic acid to recover the actinide. The filtrate is boiled, preferably in a zone of reduced pressure, to evaporate water and precipitate the alkali metal carbonate which is recovered therefrom for recycle to the combustion chamber. The filtrate, after carbonate removal, is recycled to quench additional molten salt. In a particular preferred embodiment, wherein the waste material also contains halides, the filtrate, after carbonate removal, is cooled to precip-itate an alkali metal halide which is recovered for disposal.
The filtrate then is recycled for use as the aqueous medium to quench additional molten salt.

Description

lO9iO33 BACRGROUND OF THE INVENTION
Field of the Invention This invention relates to a waste control process for the treatment of an actinide-containing organic waste. It particularly relates to a molten salt process for reducing the volume of an organic waste material contaminated with radioactive actinide elements and further provides for the separation and recovery of such radioactive actinide elements.
Prior~Art In the processing of fuel for nuclear reactors, and in the operation of such reactors, a considerable amount of waste material is generated, which is contaminated with radioactive actinide elements. It is repor$ed that the various U.S. Energy Research and Development Administration (ERDA) facilities lS generate approximately 350,000 cu.ft. of solid transuranic waste per year. Since the current costs of storing such waste are high and are likely to increase, there i8 an urgent economic incentive to reduce the volume of such waste. In addition, the high toxicity, high specific activity and long-retention in the body (half-life 300 years~ of plutonium requires s,pecial handling. Since about 1970, plutonium wastes-have been stored on a temporary basis until ultimate disposal techniques could ~ be developed. The wastes are contained in polyethylene-lined ! 55 gallon drums so as to be retrievable without external con-tamination. As much as 1600 kgms of plutonium are associated with the waste currently in interim storage.

lV91033 The majority of the waste is made up of organic combustible materials such as rags, paper, plastic and rubber.
A summary of the characteristics of the solid waste is given below.

CHARACTERISTICS OF TYPICAL SOLI~ TRANSURANIC WASTE
(Source: ERDA3 Composition (wt%) Paper 55 Rags 5 Plastic 30 (50% polyvinyl chloride and 50% polyethelene) Rubber 10 Bul~ Density (lb/cf) 7 Ash Content (%) 8 Plutonium Content tg/lb)0.014 Heating Value (Btu/lb)9,000 to 12,000 Because of the high proportion of halogenated (usually chlorinated) plastic and the danger of plutonium carryover in the particulate, a conventional incinerator is not wholly satisfactory for the combustion of this material as a means of volume reduction. Thus, more complex incinerators or special combustion methods are reguired. More particularly, in processing such waste, ideally as much volume reduction as possible is obtained with a minimum of pollution.

1~91033 Various processes have been suggested for treating different radioactive waste. None of these processes, however, have proven altogether satisfactory. British Patent No.
1,035,330 discloses a process and apparatus for treating solid radioactive wastes. The patent suggests that low level radioactive wastes be incinerated in a furnace to reduce their volume, and a multi-step filtering technique for combustion gases is proposed. The disadvantage of this process is that it requires ela~orate filters for the offgases, and further, makes no provision for recovery of radioactive actinide elements.
More particularly, since normal incineration temperatures are around 800C, it would be extremely difficult to recover the actinide elements as most of such elements form refractory oxides at temperatures in excess of around 750C. The refractory transuranic oxides are not readily amenable to conventional recovery techniques.
U. S. Pat. No. 3,479,295 suggests a method of reducinq a radioactive waste solution, obtained in the processing of nuclear fuel elements, to dryness. Broadly, the process comprises blowing an oxygen-containing gas upwardly through a bed of particles, formed by calcination of the salts in the waste solution, to fluidize the bed and feeding additional waste solution into the fluidized bed so formed. A hydrocarbon fuel also is introduced into the fluidized bed in the presence Z5 of nitrate ions at a temperature above the ignition temperature of the fuel to burn the fuel and provide the heat necessary to evaporate the solution and calcine the salts contained therein.

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1~19103;~

A disadvantage of this process is that it requires a source of nitrate ions and does not provide for the recovery of any actinide elements contained in the waste solution.
In U. S. Pat. No. 3,716,490 there is disclosed another method for the treatment of radioactive liquids. The method comprises providing a solid, fusible, partly sulfonated bituminous substance and contacting that substance with a liquid waste containing radioactive ions to ion exchange the radioactive ions with the suifonated portion of the bituminous substance. Thereafter, the bituminous substance is melted to reduce its volume and encapsulate the radioactive ions. Thus, while this patent provides a method for reducing the volume of radioactive waste, it does not provide a means for isolation and recovery of the radioactive elements.
U. S. Pat. No. 3,764,552 discloses a method for storing radioactive combustible waste material. The method comprises the steps of placing the waste material in a container provided with oxide getters selected from the group consisting of magnesium oxide, calcium oxide, barium oxide, and strontium oxide in an amount sufficient to react with sorbed water and combustion products formed by oxidation and pyrolysis of the waste material. The container then is sealed and heated to pyrolyze the waste.

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In other processes, an actinide-containing waste material is combusted and encapsulated. Examples of patents relating to such processes are U. S. Pat. Nos. 3,00~,904; 3,262,885 and 3,332,8~4.
The combustion per se of carbonaceous fuels and carbon-containing wastes in a molten alkali metal salt for various purposes is kno~n. U. S. Pat. No. 3,710,737 shows the generation of heat for external use employing a variety of carbonaceous materials. U. S. Pat. Nos. 3,567,412, 3,708,270 and 3,916,617 show the use of such techniques for the production of pyrolysis gases. In U. S. Pat. Nos.
3,77~,320 and 3,845,190, such techniques are involved, respectively, in the non-polluting disposal of explosives and of organic pesticides. In U. S. Pat. No. 3,8g9,322, valuable metals are recovered from organic scrap in a molten salt bath. None of these patents are concerned with the treatment of radioactive wastes for the isolation and retention of volatile radioactive elements.

SUMMARY OF THE INVENTION
In accordance with the present invention, there is provided a method of treating organic wastes containing at least one actinide element to reduce the volume of said waste and recover the actinide element therefrom. Broadly, the method comprises introducing an actinide-containing organic waste and a source of gaseous oxygen, such as air, into a molten salt comprising an alkali metal carbonate. The bath is maintained at a temperature of about 400C to 1000C and a pressure of from about 0.5 to 10 atmospheres to thermally 3~.0~

decompose and at least partially oxidize the actinide-containing waste. Under such conditions the volume of organic waste is substantially reduced, and combustion products are formed, which include a gaseous effluent consisting essentially of carbon dioxide and water vapor.
The gaseous effluent is vented to the atmosphere preferably through a series of filters to remove any trace amount of the actinide element which may be entrained therein, or particulate alkali metal salts which may be entrained in the gaseous effluent. The remaining combustion products of the organic waste are retained in the molten salt as will be explained more fully later. At least ~ portion of the molten salt containing the combustion products is withdrawn and mixed with an aqueous medium. The aqueous medium then is treated to remove the insoluble combustion products forming a substantially solids-free solution. The insoluble combustion products formed contain the actinide element and are leached with an inorganic acid to solubilize the actinide elements and recover them from the combustion products. The actinide elements are readily recoverable from the acid solution utilizing conventional solvent extraction or anion exchange resin techni~ues known to those versed in the art.
Preferably the alkali metal carbonate is sodium carbonate which may optionally contain from about 1 to 25 wt.% of sodium sulfate. An advantage of using sodium carbonate is that it is lower in cost, and further, when substantial portions of the carbonate are reacted with other consti-tuents of the waste, such as halides and sulfur, to form 10~ 1033 alkali metal halides and sulfates, the sodium form is more readily amenable to processing to recover the sulfur and reform the carbonate constituents for recycle in the process.
When sodium carbonate is used the preferred range of temper-ature and pressure is from about 800C to 900C and 0.8 to 1.0 atmosphere, respectively.
When the organic waste also contains sulfur and halogen constituents, the sulfur and halogens react with the alkali metal carbonate to form alkali metal sulfates and halides which are retained in the molten salt. Such wastes are typical of those which result from processing nuclear reactor fuel, for example. When such wastes are used as a feed material, there also is provided a method wherein the carbonate and sulfate component of the molten salt is recoverable for recycle.
A method also is provided for the separation of the alkali metal halide component for disposal.
In its preferred aspects, the combustion is a complete one, whether or not air, oxygen-enriched air, or fine oxygen is used as the oxygen-containing gas. Although air is gener-ally preferred, pure oxygen can be used where it is desiredto reduce the volume of gaseo~s products.

BRIEF DESCRIPTION OF THE DRAWING
The sole figure is a block diagram illustrating the method of the present invention.

DESCRIPTION OF PREFERRED EMBODIMENTS
The present invention relates to the treatment of combustible organic waste containing one or more elements of the actinide series. A significant source of such waste arises from the operation and utilization of atomic energy such as nuclear power plants. The waste resulting from the operation of such plants can be solid, liquid or gaseous.
In addition to the radioactive waste produced in the reactors involved, other wastes are developed in operation and maintenance of those installations. For example, tools, equipment and clothing that are used in close proximity to these installations may themselves become radioactive. Similarly, paper, rags and organic solvents used in such installations may also be contaminated as a consequence of such use. Further, in processing or reprocessing of the used fuel elements, waste materials are developed. The present invention is applicable to all combustible organic wastes, including such items as plastic, paper and rags. In view of the importance of having a convenient, rapid, safe, effective method of economic interest for complete treatment and reduction of volume of radioactive actinide-contaminated organic waste, the present invention will be particularly described with reference to the treatment of such waste, and particularly to the recovery of uranium and plutonium of the actinide series from such waste.
In accordance with the present method, the molten salt may consist of a single alkali metal carbonate or a mixture of alkali metal carbonates. Preferably, the molten salt also will contain a minor amount of alkali metal sulfate, since it has been found the presence of sulfate enhances the combustion ; rate of organic materials. The organic waste is combusted;
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some of the combustion products being retained in the molten salt, while others are further oxidized to CO2 and H2O.
When it is desired to effect the combustion of tne waste at a relatively low temperature, a low melting binary or a ternary mixture of alkali metal carbonates may be utilized.
For example, the ternary alkali metal carbonate eutectic consisting of 43.5, 31.5 and 25.0 mole % of carbonates of lithium, sodium and potassium, respectively, melts at 397C.
A preferred binary mixture is the sodium carbonate-potassium L0 carbonate eutectic which melts at 710C. The alkali metal sulfate utilized may consist of any of the sulfates of the foregoing alkali metals. In general, sodium sulfate is preferred because of its ready availability and low cost.
Referring to the drawing, an overall flow diagram for the method is shown. The waste, after manual sorting, is shredded in a shredder 1 and fed with air into a molten salt-containing furnace 2. Under steady state conditions, a typical composition of salt in the furnace is approximately 50 wt.% sodium carbonate, 10 wt.% sodium sulfate, 20 wt.%
sodium chloride and 20 wt.% ash. The sodium chloride results from, for example, the combustion of plastics such as poly-vinyl chloride ~its chloride content is converted to sodium chloride in the molten salt). The ash principally comprises the inorganic constituents of the waste. From processing and viscosity considerations, an upper limit for the inorganic ash concentration (insolubles) of approximately 25 wt% has been set assuming an operating temperature of about ~00C.

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The desired steady state concentrations are maintained by periodically or continuously draining a portion of the molten salt for treatment to remove the insolubles and the alkali metal halide content and further provide for recycle of the alkali metal carbonate.
In operation, the alkali metal carbonate serves as a heat transfer medium for the combustible material and also as a neutralizing agent for any acidic gaseous combustion products from the waste material such as hydrochloric acid and sulfur dioxide. The actinide-contaminated organic waste and air are fed continuously into the molten salt furnace 2 below the surface of the salt, so that all gases formed during combustion are forced to pass through the molten salt before being released into the atmosphere. Thus, in accordance with the present method, the only gaseous combustion products involved are CO2 and water vapor. The effluent gas also will contain some unreacted oxygen and nitrogen when air is used as a source of ; gaseous oxygen. The inorganic content of the waste, such as metal, is oxidized and retained in the salt principally as the ash. Another advantage of the present invention is that the temperatures of combustion are sufficiently low so that no significant amount of nitrogen oxides are formed during combustion.
In addition, the actinide elements do not form refractory oxides in the molten salt even at high temperatures, i.e., 750 to 1000C.
- The melt-ash mixture withdrawn from the furnace by way of a conduit 3 preferably is mixed with recycled solution by way of a conduit 4 to dissolve the soluble sodium salts in a vessel 5. The solution then is treated using any conven-3 tional solids liquid separation technique, such as a centrifuge 1~9 103~
or the like, to separate and remove the insoluble ash by way of a conduit 6. Advantageously, the ash is leached with an inorganic acid at 7 to solubilize the actinide constituents for recovery and for use, for examp e, as a source of nuclear reactor fuel. Suitable inorganic acids include HCl, HN03, HF and mixtures thereof.
The substantially solids-free filtrate then is treated in a series of evaporative crystallizer zones 8 and 9 at succeedingly lower temperatures to remove first a sodium carbonate or mixed sodium carbonate-sodium sulfate fraction and then a sodium chloride fraction. This phase of the method is somewhat analogous to the treatment of brines commonly practiced, for example, at Trona, California, in the treatment of the Searles Lake brine. More particularly, it has been found that it is possible to achieve a selective precipitation of salts in the mixed brine, whereby it is possible to remove a first crop of carbonate or mixed sodium æulfate-sodium carbonate crystals frequently referred to as burkeite, and subsequently, at a lower temperature to recover a second crop of crystals consisting essentially of alkali metal chlorides substantially free of carbonate or burkeite.
Examples of patents relating to the separation of sodium salts from mixtures thereof are U. S. Pat. Nos. 1,836,426 and U. S. 2,347,053. Generally, the first crop of crystals is obtained by boiling the filtrate in an evaporator crystallizer operated at a temperature of between about 60 and 170C.
After separation of this crop of crystals, the substantially crystal-free liquor is then introduced into a second evaporator crystallizer, which preferably is maintained under a vacuum, 0;~

and the liquor is cooled to a temperature of less than about43 C. Generally, a temperature of 20 to 35C is used to produce a second crop of crystals, comprising the alkali metal chlorides, substantially free of alkali metal carbonates and sulfates.
It is seen, therefore, that the present invention provides a closed cycle and produces a substantially pollutant-free offgas containing carbon dioxide and steam. In addition, there are no liquid wastes in the process, but rather two solid products for disposal, namely the ash and sodium chloride.
If desired, these can be consolidated by melting the sodium chloride, adding the ash and solidifying the mass. Alternatively, they may, of course, be kept separate for different disposal options.
The following examples are set forth to further illustrate the practice of the present invention, and are not intended to be construed as limited in scope.

The following example demonstrates the application of the present invention to a waste which simulates the character-istics of radioactive transuranic waste materials. Two different ash-melt mixtures were used in these combustion tests. The first mixture consisting of 16 wt.~ ash and 16 wt.% NaCl, lQ wt.% ~a2SO4, and 58 wt.% Na2CO3; the second mixture consisting of 20 wt.% ash, 20 wt.% NaCl, 10 wt.~
; Na2S04 and 58 wt.~ Na2CO3. These two mixtures were selected as representing the concentration extremes which would be anticipated in the actual treatment of radioactive actinide-containing waste. The composition of a typical waste was obtained from the Los Alamos Scientific Laboratory and comprised paper and plastic mixtures having essentially the same heating value, ash content, ash composition and halide 10~103;~

content found in actual radioactive-contaminated waste. The waste and air were introduced below the surface of the molten salt-ash mixture in a furnace. The offgas from the combustion was monitored for N2, 2~ C02, CO, HC and NOx to determine the completeness of combustion. Particulate samples of the cffgas also were collected to determine the particular loading and composition of the particulates. The offgas was scrubbed with aqueous sodium hydroxides to determine if any gaseous chloride escaped from the melt and passed through the particulate filter. None was found. The results of these tests are shown in Table 1.

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_ V 0:1 O . jz lV~ 33 From the foregoing table, it is seen that the nitrogen oxide emissions are less than 20 ppm, and CO and HC emissions are below the detection limits of the instruments. Further, it is seen that below about 800C the particulate emissions are within the EPA standards for particulate emissions from incinerators; i.e~, 0.1 grains/scf (0.2 gm/m3). It was found that the amount of particulate is directly related to the sodium chloride vapor pressure and mole fraction of sodium chloride in the melt. Specifically, analysis of the particulates indicates that at 800C they are essentially sodium chloride crystals.

The following example demonstrates the treatment of the melt-ash mixture to recover the sodium carbonate and sulfate for recycle to the combustor and to separate an ash and sodium chloride fraction for disposal. The melt-ash mixture was quenched in an aqueous mediu~ comprising sodium carbonate, sulfate and chloride. The aqueous medium then was filtered to remove the insoluble ash constituents. The filtrate was subjected to a first fractional crystallation at a temperature of about 106C
to precipitate a first mixed crystal crop of sodium carbonate and sodium sulfate, which was removed by filtration. The filtrate from this step was introduced into a second crystallization zone wherein sodium chloride was precipitated at a temperature of about 35C. Samples of the crystals were obtained and analyzed and the results are listed in Table ~.

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COMPOSITION OP` CRYSTALS OBTAINED IN
AQUEOUS PROCESSING TESTS
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Composition Crystallization Flow Sheet Experimental (Calculations~ Results :
1st 84% Na2CO3 80~ Na2CO3 16% Na2SO4 12% Na2SO4 8% NaCl 2nd 100% NaCl 96% NaCl 4% Na2CO3 0.2% Na2SO4 For comparison, the theoretical composition of the crystals, based on solubility calculations, also is given.
From the foregoing table it is seen that the carbonate-sulfate crystals were substantially free of sodium chloride;
i.e., less than 10%. Further, in the second crystallization it is seen that sodium chloride was obtained substantially free of carbonate and sulfate. Since the crystals were not washed, the difference between the experimental results and the theoretical is due primarily to the retention of mother liquor in the crystals. Based on weight changes, when the crystals were dried, and the composition of the solution from which the crystals were separated, it is found that nearly 80 wt% of the undried crystal weight was mother liquor in the first crystallization and about 40 wt% in the second crystallization. Thus, the use of a centrifuge instead of a filter, for example, to separate the crystals would reduce the liquor retention in the crystals, and thus even further improve the crystal purity.

10'~11~33 The following example will demonstrate the disposition of the actinide elements when treated in accordance with the present method. Four different plutonium and uranium compounds were added as solutions to separate ash-melt mixtures prepared from combustion tests. One hundred grams of crushed-ash melt mixture was placed in a 1/2 inch alumina tube, and a plutonium compound (approximately 2 ml) and corresponding uranium compound (approximately 2 ml) were placed in the center of the ash-melt mixture. The mixture was melted, and gas was bubbled through the melt to simulate the agitation present during combustion.
The test conditions are given in Table 3.
After each test the alumina tube was transferred to an analytical glove box. The alumina tube was cracked away from the solidified ash-melt mixture which was subsequently crushed to less than about 400 mesh. About 12.2 grams of the ash-melt mixture were placed in a reflux flask and 25 grams of a recycle solution (dilute sodium chloride-carbonate-sulfate solution) were added. The ash-melt mixture was refluxed with the recycled solution for two hours. The mixture then was filtered to separate the ash from the solution~
Aliquots of the reflux filtrate, the water used to wash any particulates out of the condenser and sections of the filter were analyzed to determine if any plutonium or uranium was carried out of the melt during simulated combustion. The uxanium in the various fractions was determined by a fluorometric procedure.

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'. --19--The disposition of plutonium and uranium in the various processing fractions and components is shown in Table 4. The percent of the original plutonium and uranium in the ash fraction was obtained by difference. In addition, one test was made to determine if plutonium dissolved from the ash fraction. The ash was leached with a mixture of HF-HN03 and plutonium so dissolved was determined coulometrically, the result confirming the percent calculated by difference.
From the table it is seen that g9.9% of plutonium was present in the ash fraction. Only a small percentage of the plutonium (generally less than 0.1%) was present in the reflux solution. Only a negligible amount of plutonium was carried away with the product gases, generally less than about 0.01%.
Further, it is seen that essentially no difference in the particulate evolution is observed with gas-melt superficial velocities of from about 1 to 2 ft/sec. In the actual combustion of radioactive-actinide-containing waste it is anticipated that superficial velocities of from 0.5 to about 2 ft/sec will be satisfactory.
The uranium results are similar to that of plutonium.
As with plutonium, only trace amounts of uranium were found in the heat exchanger and filter used. However, unlike plutonium, it appears as if a small fraction of the uranium may dissolve in the reflux solution. Nonetheless, by far, the majority of uranium is present in the ash.

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EXAMP~E 4 The following example is set forth to demonstrate the combustion of an actinide-containing organic waste to reduce its volume followed by the recovery of the actinide from the combustion products. Three combustion tests were conducted with plutonium-contaminated waste. The waste was a mixture of paper, paper products, plastic (a mixture of polyethylene and polyvinyl chloride), and rubber. The waste was contaminated with plutonium by adding a known volume of plutonium nitrate, sulfate, and chloride solutions to the waste. After contamination, the waste was mixed thoroughly. In two of the tests, the plutonium concentration in the waste was 9 x lO 5 grams per gram of waste, and in one test the plutonium concentration was l.l x 10 3 grams per gram, which corresponds roughly to the level of plutonium expected in low level and intermediate level actinide-containing ~ wastes, respectively.
! The waste was introduced into a molten sodium carbonate bath maintained at a temperature within the range from about 850 to 905C. The off-gas from each test was monitored and analyzed for plutonium content. By difference it was determined that about 99.9% of the plutonium was retained in the melt.
To demonstrate that the actinide ~plutonium) could be recovered from the spent salt, solidified carbonate from the foregoing test was dissolved in water and filtered. The insoluble ash (containing the plutonium) was leached with various inorganic acids. The results are shown in Table 5.

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-~3-1.0~ 10:33 From the foregoing table it i~ seen th~t greater than 90%
;, of the plutonium is recovered. Further, as a result of the foregoing examples, it was determined that a substantial reduction in the volume of organic waste was obtained. More particularly, even when the entire melt-ash mixture is withdrawn for disposal, the weight of the organic waste is reduced by a factor of 2.5, and the volume of waste is reduced by a factor of 25. However, in accordance with the particularly preferred embodiment, wherein the melt-ash mixture is treated to recover and recycle the carbonate-sulfate fraction, the weight is reduced by a factor of 5.7, and the volume of waste is reduced by a factor of 57. Thus, , the foregoing examples clearly demonstrate the efficacy and advantage of the present method for the treatment of actinide-containing waste.
While the present process has been described with regard to certain particular sources of waste, actinide elements, process conditions, temperatures, concentrations and the like, and has been illustrated, in part, with various synthetically prepared radioactive wastes, it will be readily apparent to those versed in the art that many variations thereof may be used. Accordingly, this invention is not to be limited by the illustrative and specific embodiments thereof. Rather, its scope should be determined in accordance with the following claims.

Claims (10)

THE EMBODIMENTS OF THE INVENTION IN WHICH AN EXCLUSIVE
PROPERTY OR PRIVILEGE IS CLAIMED ARE DEFINED AS FOLLOWS:

1. A method of treating an organic waste containing at least one actinide element to reduce the volume of said waste and recover the actinide element therefrom comprising:
introducing the waste and oxygen into a molten salt comprising an alkali metal carbonate bath maintained at a temperature of from 750°C to 1000°C and a pressure of from about 0.5 to 10 atmospheres to at least partially oxidize said waste to reduce the volume of said waste and form combustion products, including a gaseous effluent consisting essentially of carbon dioxide and water vapor, venting said gaseous effluent to the atmosphere, the remaining combustion products of the waste remaining in the molten salt;
withdrawing at least a portion of the molten salt containing combustion products and mixing said molten salt with an aqueous medium;
removing the insoluble combustion products from the aqueous medium to form a substantially solids-free solution, and leaching the removed insoluble combustion products with an inorganic acid to solubilize and recover the actinide elements.

2. The method of Claim 1 wherein said molten salt is maintained at a temperature of between about 800° and 900°C, a pressure of from 0.8 to 1.0 atmosphere, and consists essentially of sodium carbonate and optionally contains from about 1 to 25 wt.% sodium sulfate.

3. The method of Claim 1 wherein said organic waste also contains sulfur and halogen constituents which react with the alkali metal carbonate to form alkali metal sulfates and alkali metal halides.

4. The method of Claim 3 wherein the solids free-solution is boiled to evaporate water, whereby there is precipitated a first crop of mixed crystals comprising alkali metal carbonate and sulfates, and recovering said crystals to provide a crystal-free solution.

5. The method of Claim 4 wherein said crystal-free solution is cooled to precipitate a second crop of crystals comprising alkali-metal halides which are recovered from the solution.

6. The method of Claim 5 wherein the solution after recovery of the alkali metal halides is returned and used as the aqueous medium for mixing with additional molten salt.

7. The method of Claim 1 wherein said actinide element is selected from the group consisting of uranium and plutonium.

8. The method of Claim 1 wherein said inorganic acid is a mixture of HF and HNO3.

9. The method of Claim 1 wherein said organic waste is a radioactive waste derived from the processing of nuclear reactor fuel.

10. The method of Claim 1 wherein said inorganic acid comprises an aqueous solution of HCl.