SE517130C2 - Method and apparatus for volume reduction of radioactive oil scale waste - Google Patents

Abstract

A process for the volume reduction of radioactive oil scale, comprising: providing a de-oiled solid scale material, dissolving barium and strontium sulphates and associated radioactivity thereof, removing sulphate ion from the solution obtained by means of an anion exchange operation and separating radioactive and non-radioactive constituents, by means of cation ion exchange operation. Apparatus for such volume reduction, comprising a dissolver vessel at least one anion exchange column, at least one cation exchange column, and recovery vessels.

Description

20 25 30 517 130 u ø c u -n originally according to the needs that applied to the nuclear power industry, which uses small plants compared to those that occur in connection with oil extraction.

It is the volume of the waste and not its radioactive content that is the main problem in this respect.

Dissolution of barium sulphate in oil scale has also become a way of treating the same. There are many examples of descriptions of “in-situ” dissolution of the oil scale itself, together with associated radioactivity. EDTA (ethylenediaminetetraacetic acid) or DPTA (diethylenetriaminepentaacetic acid) are commonly used (FJ Quattrini, 3,660,287, JB Olson and PK Nolan "The Chemical Dissolution of Barium Sulphate (Barite)", Corro- '92, GH Hardy and ZI Khatib, SPE 36586, Treatment and Disposal Options for NORM Oil Field 1996). can be significantly improved if synergists are added, such as dissolution media 1972, U.S. Patent Paper No. 26, Waste, It has been found that dissolution rate- such as fluoride or oxalate anion (JM Paul, "An Improved Solvent for Sulphate Scales", Corrosion '94). Many other more exotic synergists, such as macrocyclic compounds , DN Reinhoudt, GJ Torny-Schutte and A. Van Zon, patent NO 146820, 1982).

Furthermore, scale inhibition methods have been proposed, e.g., ar, it has also been described that sulphate should be removed from seawater in order to prevent the formation of scale or the addition of scale inhibitors for the same purpose.

To the best of our knowledge, however, no method has been described for reducing the volume of radioactive oil scale, by means of which the bulk of the radioactivity is simultaneously removed from the non-radioactive portion of the scale in question. General description of the invention From the above it appears that there is a considerable potential benefit in processing oil scale in order to reduce its volume before it is deposited or handled in another way.

This is achieved with the aid of the present invention. In other words, an object of the invention is to reduce the volume of oil scale before it is deposited or handled. Another object is to remove and separate radioactive constituents in the scale from non-radioactive ones. Other objects and advantages of the invention will become apparent to one skilled in the art upon a reading of the following more detailed description.

Thus, in accordance with the present invention, it has been found that by using a method of ex-situ dissolution, a relatively uncomplicated and efficient volume reduction of oil scale waste can be achieved, which method also leads to an efficient separation of a radioactive fraction from a non-volatile oil. radioactive fraction originating from the said scale.

The non-radioactive constituents can then be disposed of or recycled without control of radioactive material. Before the present invention was made, it was thought that it would be possible to achieve sufficient volume reduction simply by burning or pyrolyzing the organic (oil) content in the scale in question, but experiments in this regard have shown that the resulting inorganic residue still occupies excessive volume.

The inorganic chemical content of the said scale has been found to be mainly barium / strontium sulphate, but there is great variation between samples, and samples from at least one oil field have been found to contain significant amounts of zinc ore. Co-precipitation of radium on barium sulphate is well known, so it is no surprise that barium / strontium sulphate in scale gives a concentration of radium present in the environment. The dissolution of the barium / strontium sulphate component in said oil scale in accordance with the present invention is advantageous because - if barium / strontium sulphate constitutes the main part of the sample, any insoluble residue may then be combined with the radioactive waste for disposal, which still leads to good overall volume reduction; or - if barium / strontium sulphate does not form the bulk of the sample, any radioactivity in the sample is still likely to co-precipitate on barium sulphate, the solid material remaining after dissolution being suitable for disposal or handling as non-radioactive material.

As can be seen from the above background to the invention, dissolution of barium sulphate in oil scale is known per se as a method of treating scale. For these in-situ applications, however, there are two major differences in relation to the present volume reduction method: - In-situ dissolution must take place quickly via direct contact with the newly formed scale with a low surface size. This places high demands on the kinetics of the resolution system. In contrast, there are various possibilities to improve the kinetics of the resolution of the “ex-situ” scale, such as grinding or other decomposition.

- The desired volume reduction process places additional demands on the dissolution chemistry, as this must be compatible with the subsequent separation of non-radioactive constituents from the radioactivity.

According to the invention, the scale solution obtained in the dissolution operation is then subjected to a separation step for separation of radium from barium, and it has been found that by means of the combined dissolution separation technique it is possible to eliminate sufficient radioactivity from the original scale in a integrated process, which makes it possible to dispose of or dispose of residual material as non-radioactive waste 10 15 20 25 30 35 51 7 13.

The separation of radioactivity from barium in the dissolved barium sulphate is a challenge, as the main radioactive component is radium, whose chemistry is very similar to that of barium. Marie Curie originally achieved such a separation (via repeated fractional crystallizations). More modern separation methods have since been reported in the literature. In particular, reference may be made to a method by which radium is separated from barium from EDTA solutions using ion exchange methods (G. Gleason, “An Improved Ion Exchange Procedure for the Separation of Barium from Radium). um ", Proc. Conf. Anal. Chem. Energy Technol., Oak Ridge, Tennessee, 1980). ion exchange column, EDTA is not a suitable eluent for the ion exchange separation and DCyTA (diaminocyclohexanthetetraacetic acid) is preferred.

These known methods for separating barium from radium have generally been developed for analytical purposes, and in particular they show certain important differences in relation to the objects of the present invention. In particular, the following applies: - Analytical methods require a completely 'pure' separation, which for laboratory purposes must be achieved by elution in a single column. The use of ion exchange columns in cascade (see below) would make it possible to successfully utilize an incomplete separation in a large scale chemical process.

- For analytical methods, the same restrictions (for cost reasons) do not apply to the use of exotic materials as in large-scale processes.

- The analytical methods have largely not been developed to deal with the pollution pattern that exists in the oil scale. lO 15 20 25 30 35. . o n u "n n o. n 0 'z _", .I .. »-. . . . . . . . .. - _ _ _. nu n n u n o I l _ _. _ u. ... ... . . . : _ _ _ -; : °. ' J. ... _ .. .... ....

. Thus, the present invention aims to eliminate the above problems by providing an integrated process for dissolving the radioactive constituents in oil scale and separating the dissolved products into non-radioactive and radioactive fractions.

Detailed Description of the Invention More specifically, a first aspect of the invention relates to a process for volume reduction of radioactive oil scale waste, which comprises: a) providing an oil-free solid scale material; b) exposing this solid scale material to a solvent capable of dissolving barium and strontium sulphates, together with associated radioactive elements, primarily radium, to form a scale solution; c) separates any undissolved solid from the scale solution; d) leading the scale solution to a separation step to effect separation of radium, and optionally and other radioactive elements, from barium and strontium, optionally other (other) non-radioactive component (s), radioactive fractions; in separate radioactive and non- and optionally, respectively, e) converts said radioactive and / or non-radioactive fraction (s) into solid form, whereby the radioactive fraction is reduced in volume and in a form suitable for disposal.

An object of the present invention is to enable the processing of as many different types of oil scale materials as possible with the aid of a single oil scale processing plant. In general, however, a scale based on barium sulphate is meant. It is probable that the processing plant will have a central stationary location, which serves many oil extraction plants, rather than that it consists of a mobile plant that seeks individual plants. However, the latter option is not excluded.

An object of the present invention is also to be able to recycle and reuse the various chemicals used in the process, this in order to avoid costs for new purchases and for minimizing the generation of secondary waste.

Thus, in the first step of the process according to the invention, an oil-free solid scale material is provided. In general, this means that the radioactive oil scale is collected and brought to a processing plant. The manner in which the oil scale is collected is completely conventional and may include mechanical removal from oilfield equipment or water jet techniques to collect a slurry. Any recovery method which does not significantly alter the chemical nature of the oil scale in question is compatible with the present invention.

Said oil scale is then subjected, if necessary, to drying and treatment to remove the oil content in the scale. The methods for removing oil are conventional and include steam distillation, solvent extraction and heat treatment (eg pyrolysis). An example of the latter method of treatment is described in U.S. Patent No. 5,909,654. It is unlikely that the oil recovered in this part of the process contains significant radioactivity, and it can therefore be handled in a conventional manner (eg combustion). The main reason why oil is removed is that you want to avoid oil contamination in later procedure steps, especially if ion exchange is used.

However, the first step of the process of the present invention also comprises an in-situ solution for collecting said oil scale. The solution obtained according to this methodology can then, for example, be fed directly to a pyrolysis unit to provide a 10 l 25 20 25 30 35 517 130 n. «N o - n oc '' '' '. . . . .o 8 oil-free solid scale material for treatment according to the rest of the procedure. by the process, this solid scale material is preferably provided in particulate form. This can be achieved in step a) by conventional atomization or grinding, whereby the kinetics of the subsequent dissolution are improved. Whether this step is necessary or preferred depends on the particle size of the incoming solid oil scale and the function or power of the equipment used for the subsequent dissolution, but preferably the average particle size should be reduced to a diameter of 50 μm or less.

The oil-free solid scale material is then subjected to a dissolution step, which is usually performed by placing the oil-free scale in a dissolution vessel and contacting it with a dissolution solution.

This dissolution solution must have a composition suitable for dissolving barium and strontium sulphates and associated radioactive elements, primarily radium, and it must be compatible with the subsequent steps of separating the radioactivity. Exotic or expensive chemicals are not desirable for this purpose. Preferably, an alkaline aqueous solution containing a complexing agent is used. Examples of complexing agents are EDTA (ethylenediaminetetraacetic acid), DPTA (diethylenetriaminepentaacetic acid) and DCyTA (diaminocyclohexane tetraacetic acid), where EDTA is especially preferred for process reasons and also for the reason that it is inexpensive and readily available.

The concentration of the dissolving component in the solution should be the minimum concentration that is compatible with dissolution, which usually (but not exclusively) means a value in the range 0.01 - 0.5M, preferably 0.01 - 0.1M. The reason for striving to minimize the concentration is to thereby reduce the cost of chemicals, facilitate the subsequent separation of radioactive and non-radioactive constituents and facilitate the recovery of chemicals for recycling. circulation.

The pH value of the solution must be adjusted to achieve the required dissolution conditions, which usually but not exclusively mean a value between pH = 9 and pH = II. The pH value of the dissolution solution can be adjusted with any suitable alkali, but sodium hydroxide is probably the cheapest and easiest choice. Other components may be added if this is necessary to accelerate the solution, as described in the prior art, but such components may interfere with the subsequent separation reactions of the solution and should be avoided if possible. An increase in the dissolution temperature (from room temperature up to 100 ° C) may be advantageous under certain conditions.

Improved or completed resolution can be achieved via special design of the dissolution vessel and equipment. In addition to the particle size reduction that has been discussed above, certain other methods are known per se when it comes to increasing the efficiency of dissolution reactions. Examples of such are countercurrent contact, where the solid material passes through the dissolution vessel in a direction opposite to that of the solution; thereby the outgoing undissolved solid material comes into contact with the purest solution. The contact between the solid material and the solution can also be improved with the help of a “high shear” mixer.

Any undissolved solid material from the dissolution reaction vessel is separated from the dissolution solution in a conventional manner, such as via sedimentation and filtration. The solid is rinsed with clean water (if required) and divided, according to its specificity for radioactive electrical radioactivity (in Becquerel per gram), by non-radioactive handling. As has been mentioned earlier- 10 15 20 25 30 35 u,. nnu 'I 0' u .. n .. .- .. »nv 517 130: _";'s "lO re it is probable that the undissolved solid material is either a substantial proportion of the original oil scale and non- radioactive material or a very small proportion of the original solid material and radioactive material.

The solution is then passed to a separation step for the recovery of the non-radioactive components (especially strontium and barium) and for the recovery of the small amount of radioactive components in question, preferably in the form of solid material for conditioning and disposal or other waste management.

Ion exchange is the best way to achieve the separation between the non-radioactive and the radioactive constituents in this step, but other methods could also be possible. In particular, solvent extraction methodology could be designed to be comparable to the ion exchange separations described herein.

For ion exchange separation, the optimal procedure means that the combined radioactive (radium, (Sr, Ba, absorbed on at least one cation exchange column and that be- etc) and non-radioactive etc) constituents are then separated by differential elution. Before the solution is led to the said cation exchange column (s), however, the chemistry of the solution should be adjusted so that the components of the solution are adsorbed in the desired manner by the cation exchange mass. Such adjustments, which are known per se (see G. Gleason, supra), involve the removal of sulfate ion and its replacement with a suitable alternative, preferably chloride. This can be achieved by passing the solution through an anion exchange mass, in particular in chloride form.

The pH of the solution is also adjusted as desired, the UPP 'solution solution can then be passed through, preferably to a value in the range 5 - 7. cation exchange mass (s), the dissolved radioactive and non-radioactive cationic constituents being absorbed by the ion exchange column . Separation and recovery of the radioactive and non-radioactive constituents can then be achieved via the passage of an elution solution through the ion exchange column. The especially preferred embodiment involves the use of an elution solution which has a chemical composition which is the same as or equivalent to that of the solution solution but has a different pH value. However, alternatives are possible, such as various complexing agents, such as the use of EDTA in the dissolution solution and DCyTA in the elution solution. When the constituents pass downwards in the ion exchange column, under the influence of the elution solution, they will separate and therefore come out of the column at different times. The solution coming out of the ion exchange column is divided into separate radioactive and non-radioactive fractions. Correct collection or division of fractions can be facilitated by equipment that continuously analyzes the radioactivity and the chemical composition of the starting solution.

If perfect separation cannot be achieved in a single passage through the ion exchange column, the partial separation can be used to achieve a good effect by "cascading" ion exchange columns, ie feeding radium-depleted barium and barium-depleted radium to new ion exchange separation columns. It is also possible to design the ion exchange conditions so as to concentrate the radius in a series of bands or “pulses” of radioactivity which move downwards in the ion exchange column, interrupted by intermediate bands of non-radioactive material.

For final efficient separation of the fractions and disposal or use of these, the radioactive fraction and / or the non-radioactive fraction may be converted into solid form (s).

This is preferably achieved by means of one or more precipitation operations, which can be carried out in accordance with a technique known per se, for example via sulphate precipitation (s) with sulphate ions. The resulting radioactive and / or non-radioactive solid material can then be used or disposed of in any suitable manner.

Once radioactive and non-radioactive cationic constituents have been recovered, solution constituents (eg complexing agents) can be recovered, for example via appropriate pH adjustment of the effluent solutions to effect precipitation. The filtered complexing agent can be recycled. Residual saline solutions can then be discarded in the form of effluent, but if it is absolutely necessary to achieve an operation without effluent, the saline solutions can be reconstituted to acid and alkali via appropriate electrochemical processes.

According to a second aspect of the invention, there is provided a simple and efficient apparatus for carrying out a method of the type described above. This apparatus comprises: A) a solvent vessel for a solvent capable of dissolving barium and strontium sulphates together with associated radioactive elements, primarily radium, from said scale; B) at least one cation exchange column intended to be fed with the solution obtained in said dissolution vessel, for ion exchange separation of radium, and any other radioactive elements, from barium and strontium and any other non-radioactive ingredient (s) , present in said solution from the dissolution vessel; and C) recovery vessels for separated radioactive or non-radioactive fractions from said cation exchange column (s).

With respect to preferred embodiments of such an apparatus and operation thereof, reference is made primarily to the preferred embodiments of the method which have been described above. However, some advantageous apparatus constructions may be summarized as follows: the apparatus preferably also comprises an anion exchange column, prior to said cation exchange column (s), exchanging anions in the solution from the dissolution vessel before being passed to the cation exchange column ( columns).

Another preferred embodiment of the apparatus relates to an apparatus in which said cation exchange columns are represented by a plurality of cascaded cation exchange columns.

Furthermore, the apparatus preferably also comprises a de-oiling unit, which is arranged before the dissolution vessel, for removing the oil content of the radioactive oil-scale waste from which one is starting.

To effect the conversion of radioactive (a) and / or non-radioactive (a) fraction (s) into solid material, the recovery vessels are preferably provided with pH adjusting means for precipitating at least the radioactive fraction in the form of a solid radioactive waste. .

The apparatus also preferably comprises a grinding unit, before the dissolution vessel, for disintegrating solid oil scale into a particulate material, preferably to a particle size of at most 50 μm.

Finally, reference may be made to an apparatus which comprises radioactivity assay means in conjunction with the cation exchange column (s), which enables the correct collection of fractions.

Drawings The drawings show the following: Fig. 1 shows a flow chart of an embodiment of the method according to the present invention; Fig. 2 shows the cascade principle when using a plurality of cation exchange columns for the separation step; Fig. 3 shows a flow chart of an embodiment of an apparatus according to the present invention; Fig. 4 shows dissolution of activity and weight in oil scale samples as described below in Example 1; Figure 5 shows DCyTA ion exchange as described below in Example 2; and Fig. 6 shows ion exchange separation with EDTA as described below in Example 3. More specifically, the flow chart in Fig. 1 shows how oil scale is initially received and subjected to an oil removal operation. The oil-free scale is then dissolved as previously described and the resulting solution is passed, after adjustment of the chemistry, to a separation step in accordance with the invention, precipitated "barium" sulphate is obtained in a form suitable for approved disposal or by separated, tering and the precipitated “radium” sulphate is volume-reduced and is present in a form suitable for the deposition of materials with low radioactivity.

Furthermore, it is also shown how the solid material separated from the dissolution step can be handled in the form of a radioactive or non-radioactive residue.

Fig. 2 shows how an input solution is subjected to a first separation step in a cation exchange mass, whereby the barium-rich and radium-rich fractions from this first separation are then passed for separation into new cation exchange masses, etc. Solid barium sulphate and radium sulphate are recovered separately for handling, that radium- and barium-rich fractions, as shown, are recycled to the first ion exchange mass.

The flow chart in Fig. 3 shows a heat treatment unit for producing an oil-free scale material; followed by a dissolution vessel for the dissolution step; an anion exchange column for adjusting the chemistry of the solution obtained in the dissolution vessel; a cation exchange column for the separation step; as well as recovery vessels for the radioactive and non-radioactive fractions, respectively.

The operation of such an apparatus should be clear from the above procedure description and need not be repeated here. Examples The present invention will now be described by the following examples: Example 1 - Dissolution Duplicate oil scale samples from two sources ("A" and "B") were heat treated by pyrolysis of organic material and ground separately to <50 μm particle size and homogenized using a Retsch (S100) centrifugal ball mill with reversing mechanism.Two 5 g portions of each were carefully weighed and subjected to gamma spectrometric measurement to determine Ra-226 activity. was measured in the same geometry as a 50 Bq standard made from 5 g of the natural mineral clinoptilolite nailed with a water-based certified Ra-226 reference standard.

Dissolution tests with several contacts between the materials were performed on 5 g portions of the ground samples. According to this technique, the oil scale was contacted with a chemical solution, after which it was filtered and then contacted with a fresh batch of the solution. The number of contacts was registered and the total resolution of activity and weight was measured. The ground oil scale sample was added to 300 cm 3 of fresh 0.1M EDTA solution at pH 9 (Na form, pH adjusted with NaOH). This was maintained at 50 at 5 ° C in a 600 cm 3 glass beaker on a hot plate. Said oil scale and chemical solution were mixed vigorously using a laboratory type high shear mixer (Ystral GmBH type X1020) for 30 minutes. The resulting mixture was centrifuged twice at 4500 rpm for 5 minutes. The supernatant of the dissolution contact liquid was decanted off and filtered under vacuum through a 0.2 μm glass microfiber filter. The filtered solvents were subjected to 2 hours of gamma spectrometric measurement with quantification against a 100 Bq Ra-226 standard with equivalent geometry made from deionized water nailed with the certified Ra-226 reference standard. 10 15 20 25 30 35 517 130 16 The average loss of non-transferable solid material per filtration of the solution contact liquids was 9 mg, which was determined by carefully weighing dried filters after filtration in the form of duplicate samples. The final treated solid was carefully weighed and then brought to 5 g with clinoptilolite at the correct count geometry. These were subjected to 12 hours of gamma spectrometric measurement.

The results in Pig 4 show the extent of the solids dissolution and radioactivity in the two samples. In the case of sample “A”, the proportion of activity removed by dissolution was 99.3% and 99.7% in the second sample. In the case of sample “B”, where there is a significant proportion of undissolved solid material left, the radioactivity of this remaining solid material is below the clearance limit.

Example 2 - Ion exchange separation with DCYTA Column preparation A 75 cm3 cation exchange column in sodium form was prepared from Amberlite IR-120 pulp (capacity 1.9 meq x cma).

The total cation capacity of the column was approximately four times the theoretical requirement for the barium content of the solution, thus ensuring sufficient available capacity for radium / barium separation in the presence of other metal ions in the solvent.

Removal of sulphate and pH change in the solution solution A 75 cma anion exchange column was prepared using a weakly basic anion exchange mass, Ionac 365, capacity 2 meq cm%. Assuming that up to 4 g of barium sulfate potentially dissolves in the dissolution liquid, the total capacity of the column meant an excess of approximately 75%. The column was converted to chloride form from hydroxide form by passing 5 bv (bed volumes) of 2M sodium chloride solution followed by 2 bv deionized water as rinsing liquid through the column. 10 15 20 25 30 35 517 130 a »~. n u q 17 The first contact solvent (300 cm 3) from an oil scale solution of the type described in Example 1 was charged to the column at a flow rate of l1.5 cnfmind followed by rinsing with 4 bv of deionized water.

No precipitation was observed in the column, and the fact that sulfate had been successfully removed was verified using Merckoquant Sulfate-Test strips, detection limit <200 mgld. The total column effluent was combined and a 300 cm? aliquot was counted so as to ensure that the bulk of the activity had passed through the resin. The combined solution was nailed with 1000 Bq Ra-226 (from a certified standard in HCl) and again subjected to a 300 cm3 aliquot for measurement. The solution was nailed with extra radium in order to facilitate the subsequent cation exchange experiments. Half of the nailed solution (pH 8.5) was reduced to pH 7 for ion exchange separation.

Charge and Separation The solvent adjusted for pH 7 nailed with additional Ra-226 (as described above) was charged to a 75 cm 3 cation exchange column in sodium form (pH 7) prepared in the manner described above. After the loading of solvent, the column was rinsed with 2 bv of deionized water followed by another 1 bv of deionized water. 13 bv of 0.05M DCyTA at pH 8.5 was passed through the barium elution column with column effluent fractionation for measurement and qualitative 10 bv and 13 bv. Thereafter, 5 bv of 0.05M EDTA at pH 8.5 was passed by cobarium analysis after 5 bv, the radon for elution followed by rinsing with 2 bv of deionized water. As before, each fraction was subjected to measurement and qualitative barium analysis in order to save radioactivity and the presence of barium. The majority of barium was recovered in the first 10 bed volumes of DCyTA effluent. In particular, the first five bed volumes of effluent contained barium but not radium.

The radium activity recovery is shown in Fig. 5. 10 3 20 20 5 517 130 n: unt 18 Example 3 - Ion exchange separation with EDTA The first oil scale solution contact liquids from and B (5) are removed by anion exchange as described above. Samples A (5) were combined and subjected to sulfate-combined anion exchange effluent adjusted to pH 7 with 2M HCl. The whole solution (unspiked, combined Ra-226 activity ~ 500 Bq) was charged to a 75 cm 3 cation exchange column in sodium form (pH 7), tet. A total of 12 bv prepared in the method described above (ie 900 ml) Fractions were taken after 4 bv and each subsequent 2 bv for solution was charged. obtaining an indication of the point (s) at which radium and barium breakthroughs occurred. After the addition of the combined solvents, the column was rinsed with 2 bv of deionized water. Then 2 x 5 bv (ie a total of 10 bv) of 0.05M EDTA at pH 8.5 were passed through the column for elution of radium and any remaining barium, followed by rinsing with 2 ~ bv deionized water. As before, each fraction was subjected to counting and qualitative barium analysis in order to track the radioactivity and presence of barium. The radioactivity recovered in each fraction is shown in Fig. 6. The initial penetration of barium from the ion exchanger was completely radium free, and in reality very large quantities of radium depleted barium were eluted from the column. Using the elution step, barium depleted radium was recovered.

Claims (25)

10 15 20 25 30 35 517 130 19 PATENT REQUIREMENTS A process for reducing the volume of radioactive oil scale waste, comprising: a) b) d) e) providing an oil-free solid scale material; exposes this solid scale material to a solvent capable of dissolving barium and strontium sulfates, together with associated radioactive elements, primarily radium, to form a scale solution; separates any undissolved solid material from the scale solution; passing the scale solution through an anion exchange mass to remove sulfate ion therefrom; the solution depleted in sulphate ion leads from the anion exchange mass through a cation exchange mass to effect the separation of radium, and possibly other radioactive elements, from barium and strontium, and possibly another (other) non-radioactive component (s), in separate radioactive and non-radioactive fractions; and optionally converting said radioactive and / or non-radioactive fraction (s) to solid form; whereby the radioactive fraction from the process is reduced in volume and in a form suitable for disposal. A process according to claim 1, wherein, in step a), oil is removed from collected oil scale by an operation selected from steam distillation, solvent extraction and heat treatment. A method according to any one of the preceding claims, wherein, in step a), the solid scale material is provided in particulate form, preferably by grinding, and in particular with a particle size of at most 50 μm. 10 15 20 25 30 35 517 130 20 A process according to any one of the preceding claims, wherein, in step b), the solvent is an alkaline solution containing a complexing agent. The method of claim 4, wherein the complexing agent is selected from EDTA, DPTA and DcyTA. A method according to any one of claims 4 and 5, wherein the concentration of the complexing agent is in the range 0.01-0.5 M, preferably 0.01-0.1 M. A process according to any one of claims 4-6, wherein the pH value of the alkaline solution is within 9-11. A process according to any one of the preceding claims, wherein, in step c), the separation is effected by sedimentation and filtration. A method according to any one of the preceding claims, wherein said step d) is performed by means of an anion exchange column. A method according to any one of the preceding claims, wherein the anion exchange mass is in chloride form. Process according to any one of the preceding claims, wherein step e) is carried out by means of at least one cation exchange column. Process according to any one of the preceding claims, wherein step e) is carried out at a pH in the range 5-7. A process according to any one of the preceding claims, wherein the radioactive and non-radioactive fractions are recovered by passing an elution solution through the ion exchange mass in question. The method of claim 13, wherein the elution solution is based on a complexing agent other than that used for the dissolution operation in step b). A process according to claim 13, wherein the elution solution is based on the same complexing agent used in the dissolution operation in step b), the elution being carried out at a different pH value. A method according to any one of the preceding claims, wherein step e) is carried out by means of cascade-coupled cation exchange columns so that the radium depleted barium and the barium depleted radium, respectively, are fed to further separate cation exchange columns. Process according to any one of the preceding claims, wherein, in step f), the conversion is carried out by means of a precipitation operation, preferably with sulphate precipitate (s). A process according to any one of the preceding claims, wherein dissolved component (s), such as complexing agents, are recovered after said separation of radioactive and non-radioactive fractions. A method according to claim 18, wherein the recovery is accomplished by precipitation. An apparatus for reducing the volume of radioactive oil scale waste, comprising: A) a solvent vessel for a solvent capable of dissolving barium and strontium sulfates, together with associated radioactive elements, primarily radium, from said scale; B) an anion exchange column intended to be fed with the solution obtained in the dissolution vessel, for removing sulphate ion therefrom; C) at least one cation exchange column intended to be fed with the solution from the anion exchange column, for ion exchange separation of radium, and optionally other radioactive elements, from barium and strontium and any other non-radioactive component (s) present in the sulfate ion depleted solution ; and D) recovery vessels for separated radioactive and non-radioactive fractions, respectively, from said cation exchange column (s). The apparatus of claim 20, wherein said cation exchange column (s) is represented by a plurality of cascaded cation exchange columns. Apparatus according to any one of claims 20-21, which comprises an oiling unit, arranged before the dissolving vessel, for removing oil from the starting radioactive oil scale waste. Apparatus according to any one of claims 20-22, wherein the recovery vessels are provided with pH regulating means for precipitating at least the radioactive fraction in the form of a solid radioactive waste. Apparatus according to any one of claims 20-23, which comprises a grinding unit, before the dissolution vessel, for reducing the solid oil scale to a particulate material, preferably to a particle size of at most 50 μm. An apparatus according to any one of claims 20-24, which comprises radioactivity assay means in connection with said cation exchange column (s).