WO2002013202A1

WO2002013202A1 - Oil scale volume reduction

Info

Publication number: WO2002013202A1
Application number: PCT/SE2001/001723
Authority: WO
Inventors: Maria Lindberg; David Bradbury; George Richard Elder
Original assignee: Studsvik Radwaste Ab
Priority date: 2000-08-10
Filing date: 2001-08-09
Publication date: 2002-02-14
Also published as: SE0002869D0; SE0002869L; SE517130C2; AU2001280375A1

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

OIL SCALE VOLUME REDUCTION

Technical Field

The present invention relates to the field of radioactive oil scale waste. More specifically, it relates to a process and an apparatus for the volume reduction of such oil scale so as to reduce handling and storage thereof, as well as to facilitate disposal of the radioactivity therein.

Background of the invention Radioactive oil scale arises in the oil extraction industry. It contains principally salts of alkaline earth metals, and it is found on pipework and equipment used in the oil extraction process. The radioactive content of the scale is of natural origin, consisting principally of radium isotopes (Ra-226 and Ra-228) but with a smaller quantity of thorium and possibly other isotopes . The scale is not highly radioactive, but sufficiently so to cause concern. Today it is generally treated by water jetting to remove it from the equipment, and is either stored or returned to the environment.

Facilities exist in the United Kingdom and elsewhere for disposal of low level radioactive waste. Conditioning and packaging followed by disposal in these facilities would probably be an appropriate disposal route for the radioactive content of oil scale. However, there is a considerable cost associated with such disposal, and (what may be of greater concern) the potential volume of scale to be disposed of might ultimately be too great for the waste disposal sites to handle conveniently. The sites were designed for the needs of the nuclear industry, which employs small facilities in comparison with oil extraction. It is the volume of the waste, not its radioactive content that is the main problem in this regard .

Dissolution of barium sulphate in oil scale has also been established as a means of treating it. There are many examples described directed towards "in-situ" dissolution of the oil scale itself, together with its associated radioactivity. Typically EDTA (ethylene diamine tetra-acetic acid) or DPTA (diethylene triamine penta-acetic acid) are used as the dissolving media (FJ Quattrini, US Patent 3,660,287, 1972, JB Olson and PK Nolan "The Chemical Dissolution of Barium Sulphate (Barite) " Corrosion '92, paper no 26. G.H. Hardy and Z.I. Khatib, SPE 36586, Treatment and Disposal Options for NORM Oil Field Waste, 1996) . It has been found that there is a significant enhancement of the dissolution rate when synergists are added, such as fluoride or oxalate anion (JM Paul, "An Improved Solvent for Sulphate Scales",

Corrosion '94). Many other more exotic synergists, such as macrocycles, have also been described (e.g. F DeJong, DN Reinhoudt, GJ Torny-schutte and A Van Zon, Patent NO 146820, 1982) . Furthermore, scale inhibition programs have been proposed, e.g. by removing sulphate from seawater so as to prevent formation of scale or by adding scale inhibitors for the same purpose .

A process for decontamination of radioactive materials is disclosed in US 5,322,644 but according to this prior art the contaminated material contains actinides, such as uranium and trans-uranic elements, as major radioactive constituents. This necessitates the use of a carbonate solution to dissolve the contaminants in the material. That is, the starting material, and hereby the problem to be solved by said US patent, is completely different from the starting material of the present invention. Furthermore, US 5,322,644 does not disclose any separation of the type used in the present process.

However, to the best of our knowledge there has not been disclosed any process for the volume reduction of radioactive oil scale by means of which the major radioactivity is at the same time removed from the non- radioactive part of the scale.

General description of the invention

From the above-mentioned, it can be gathered that there is a considerable potential benefit in processing the oil scale to reduce its volume before disposal.

This is accomplished by means of the present invention. In other words, one object of the invention is to reduce the volume of oil scale before disposal thereof. Another object is to remove and separate radioactive constituents of the scale from the non- radioactive ones. Other objects of and advantages with the invention should be apparent to a person skilled in the art from the detailed description thereof below. Thus, in accordance with the present invention it has been found that by using an ex-situ dissolution methodology a rather non-complicated as well as efficient volume reduction of an oil scale waste can be achieved, which also efficiently separates a radioactive fraction from a no -radioactive fraction from said scale.

The non-radioactive constituents could then be disposed of or recycled without radioactive material controls. Initially it was thought that sufficient volume reduction could be achieved just by burning or pyrolyzing the organic (oil) content of the scale, but tests on this have demonstrated that the resulting inorganic residue still occupies too great a volume. The inorganic chemical make-up of this scale has been found to be principally barium/strontium sulphate, but there is a variability between samples, and at least one oilfield sample has been found to contain significant quantities of zinc blende. The co-precipitation of radium on barium sulphate is well known, so it is no surprise that barium/strontium sulphate in the scale concentrates environmental radioactive radium.

The dissolution of the barium/strontium sulphate constituent of the oil scale in accordance with the present invention is advantageous because

- if barium/strontium sulphate constitutes the majority of the sample, then any non-soluble residue can be combined with the radioactive waste for disposal, still leading to good overall volume reduction; or - if barium/strontium sulphate does not constitute the majority of the sample, it is still likely that any radioactivity in the sample is co-precipitated on barium sulphate, in which case the residual solid after dissolution will be suitable for disposal as non- radioactive material.

As can be seen from the Background section above, dissolution of barium sulphate in oil scale is known per se as a means of treating scale. However, in these in- situ applications there are two conditions that are different in comparison with the present volume reduction process :

- In-situ dissolution must take place quickly by direct contact with the low surface area "asformed" scale. This places high demands on the kinetics of the dissolving system. By contrast there are various options for improving kinetics of dissolution of "ex-situ" scales, such as milling or grinding. - The desired volume reduction process places extra demands on the dissolution chemistry because it 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 stage to separate radium from barium, and it has been found that by the combined dissolution - separation technique, sufficient radioactivity can be eliminated from the original scale in an integrated process, to allow the remaining materials to be disposed of as non-radioactive waste.

The separation of radioactivity from barium in the dissolved barium sulphate represents a challenge, because the principle radioactive constituent is radium, which has chemistry very similar to barium. Marie Curie originally achieved this separation (by repeated fractional crystallisation) . More modern methods of separation have since been reported in the literature. In particular, a method has been reported which separates radium from barium out of EDTA solutions using ion exchange methods (G. Gleason, "An Improved Ion Exchange Procedure for the Separation of Barium from Radium" , Proc Conf Anal Chem Energy Technol, Oak Ridge Tennessee, 1980) . In this paper it is suggested that, although EDTA solutions can be used to "load" the combined barium and radioactivity on to the ion exchange column, EDTA is not a suitable eluent for the ion exchange separation, and that DCyTA (diamino-cyclohexane tetra-acetic acid) is preferable.

These prior art methods of separating barium from radium have generally been developed for analytical purposes, and in particular have certain crucial differences from the objectives of the present invention.

In particular:

- Analytical methods require a completely "clean" separation, which, for laboratory convenience, must be achieved in a single column elution. The use of ion exchange columns in a cascade (see later) would allow the successful utilisation of an incomplete separation in a large-scale chemical process.

- Analytical methods do not have the same restrictions (due to cost) on the use of exotic materials as large- scale processes.

- Most analytical methods have not been developed to deal with the pattern of impurities present in oil scale.

Accordingly, the present invention seeks to overcome these problems by providing an integrated process for the dissolution of the radioactive constituents of the oil scale and the separation of the dissolved products into non-radioactive and radioactive fractions.

Detailed Description at the Invention

More specifically, according to a first aspect of the invention, a process is provided for the volume reduction of radioactive oil scale waste, which comprises : a) providing a de-oiled solid scale material; b) subjecting said solid scale material to a solvent having the ability of dissolving barium and strontium sulphates, together with associated radioactive species, primarily radium, therefrom so as to form a scale solution; c) separating any undissolved solid from said scale solution; d) passing the scale solution through an anion exchange resin so as to remove sulphate ion therefrom; e) passing the sulphate-ion depleted solution from the anion exchange resin through a cation exchange resin so as to accomplish separation of radium, and optionally other radioactive species, from barium and strontium, and optionally other non-radioactive constituent (s) , into separate radioactive and non-radioactive fractions, respectively; and optionally

f) converting said radioactive and/or non- radioactive fractio (s) into solid form(s), the radioactive fraction being volume reduced and in a form suitable for disposal.

It is a purpose of the present invention that as many different types of oil scale feed as possible can be processed by a single oil scale processing plant. Generally, however, a scale based on barium sulphate is referred to. Furthermore, the major radioactive species is radium. It is probable that the processing plant would have a central fixed location which serves many oil extraction facilities, rather than being a mobile plant which visits individual facilities. The latter option is not, however, precluded.

It is also a purpose of the present invention that wherever possible chemicals used in the process can be recycled and re-used to avoid the cost of re-purchasing and to minimise the generation of secondary waste.

In the first stage of the process claimed, a de- oiled solid scale material is thus provided. Generally this means that radioactive oil scale is collected and brought to a processing plant. The means by which oil scale is collected are entirely conventional and could include mechanical removal from oilfield equipment or water jetting techniques to collect a slurry. Any method of recovery that does not significantly alter the chemical nature of the oil scale is compatible with the present invention.

After receipt the oil scale is, if necessary, dried and treated to remove the oil content of the scale. The methods for removing the oil are conventional and include steam distillation, solvent extraction and heat treatment (e.g. pyrolysis) . An example of the latter method of treatment is described in US Patent 5,909,654. Any oil that is recovered in this part of the process is unlikely to contain significant radioactivity and can be disposed of by conventional means (e.g. incineration) . The main reason for removing oil is the avoidance of oil contamination in later processing stages, particularly if ion exchange is being used.

The first stage of the process according to the present invention, however, also encompasses an in-situ dissolution approach to collect the oil scale. The solution obtained in such an approach could then, for instance, be fed directly to a pyrolysis unit to create a de-oiled solid scale material for treatment through the rest of the process.

In step a) of the process said solid scale material is preferably provided in particulate form. This can be accomplished by conventional grinding or milling, to enhance the kinetics of the subsequent dissolution. The necessity or otherwise of this step is dependent on the particle size of the input oil scale solid and the performance of the subsequent dissolution equipment, but preferably the average particle size should be reduced to a diameter of 50 μm or less.

The de-oiled solid scale material is then subjected to a dissolution stage, which is generally performed by placing the prepared de-oiled scale in a dissolver vessel and contacting it with a dissolving solution.

The dissolution solution must contain a composition which is suitable for dissolving barium and strontium sulphates and their associated radioactive species, primarily radium, and must be compatible with the later process stages of radioactivity separation. Exotic or expensive chemicals are undesirable for this purpose. Preferably, an alkaline aqueous solution containing a chelating agent is used. Examples of chelating agents are EDTA (ethylene diamine tetra-acetic acid) , DPTA (diethylene triamine penta-acetic acid) and DCyTA (diamino-cyclohexane tetra-acetic acid) , EDTA being particularly preferable for process reasons and also for being inexpensive and readily available. Most preferably the chelating agent referred to is the only chemical added to the alkaline solution. For example, this means that the addition of carbonate is generally not accepted. In other words step b) is preferably performed in the absence of carbonate ion.

The concentration of the dissolving component in the solution should be the minimum compatible with dissolution, typically (but not exclusively) in the range of 0.01 - 0.5 M, preferably 0.01 - 0.1 M. The reasons for minimising concentration are the reduction of chemical costs, facilitation of later separation of radioactive and non-radioactive constituents and ease of recovery of chemicals for recycle. The pH of the dissolving solution must be adjusted to achieve the required dissolving conditions, typically but not exclusively between pH=9 and pH=ll. The dissolving solution pH can be adjusted with any suitable alkali, but sodium hydroxide is likely to be the cheapest and most convenient option. Other components can be added if necessary to accelerate dissolution, as described in the prior art, but such components may interfere with the subsequent solution separation reactions and should be avoided if possible. Increasing the dissolution temperature (from room temperature towards 100°C) can be beneficial in some circumstances.

Completion of the dissolution may be aided by the detailed design of the dissolver vessel and equipment. In addition to the particle size reduction referred to earlier, certain other methods are well known in the art for increasing the effectiveness of dissolution reactions. Examples are counter-current contact, in which the solid passes through the dissolution vessel in the opposite sense to the solution - thus the emerging undissolved solid is contacting the cleanest solution. Contact of the solid and solution can also be enhanced by a "high-shear" mixer.

Any undissolved solid from the dissolution reaction vessel is separated from the dissolution solution by conventional means, such as settling and filtration. The solid is rinsed with clean water (if necessary) and sentenced, according to its radioactivity specific activity (in Becquerels per gram) , to radioactive or non- radioactive disposal. As explained earlier, the undissolved solid is likely to be either a significant proportion of the original oil scale and non-radioactive, or a very small proportion of the original solid and radioactive.

The scale solution is then subjected to an anion exchange resin so as to remove sulphate ion therefrom. Preferably said anion exchange resin is in the chloride form, the chloride ions replacing the sulphate ions in the solution. By performing such an anion exchange operation, which can be accomplished by means of at least one anion exchange column, it has been found that the solution obtained from the dissolution of the barium sulphate, with its associated radium, can be properly adsorbed and eluted in the subsequent cation exchange operation. Further adjustments, if desired, of the chemistry of the solution can be made in accordance with prior art (see. G. Gleason, supra) . The pH of the solution subjected to the anion exchange resin is generally controlled or adjusted so as to be above 7 and below 11.

The sulphate-ion depleted solution is then passed to a separation stage to recover the non-radioactive components (particularly strontium and barium) and to recover the small amount of radioactive components, preferably in the form of solid for conditioning and disposal. Said separation is accomplished by using a cation exchange resin.

For the cation exchange separation, the optimum procedure is to absorb the combined radioactive (radium, etc.) and non-radioactive (Sr, Ba etc) constituents onto at least one cation exchange column. The separation of the constituents is then made by differential elution.

The solution pH is also properly adjusted, preferably to a value within the range of 5-7, before being passed through the cation exchange resin (s), whereupon the dissolved radioactive and non-radioactive cationic constituents are absorbed by the ion exchange resin (s) .

Separation and recovery of the radioactive and non- radioactive constituents can now be achieved by passage of an eluting solution through the ion exchange column.

The most preferred embodiment is to use an eluting solution that has a chemical composition equivalent to the dissolving solution, but at a slightly different pH. However, alternatives are possible, such as different chelating agents, e.g. the use of EDTA in the dissolving solution and DCyTA in the eluting solution. As the constituents pass down the ion exchange column, under the influence of the eluting solution, they become separated, and therefore pass out of the column at different times. The output solution from the ion exchange column is diverted to collect radioactive and non-radioactive fractions separately. The correct collection of fractions may be facilitated by equipment which continuously analyses the radioactivity and chemical composition of the output solution.

If perfect separation cannot be achieved on a single pass through the cation exchange column, the partial separation can be used to good effect by "cascading" ion exchange columns, i.e. feeding radium-depleted barium and barium-depleted radium to further ion exchange separation columns. It is also possible that the ion exchange conditions can be designed to concentrate the radium into a series of bands or "pulses" of radioactivity going down the ion exchange column, interspersed with bands of non- radioactive material.

For final effective separation of the fractions and disposal or use thereof, the radioactive fraction and/or the non-radioactive fraction is (are) optionally converted into solid form(s) .Preferably this is accomplished by means of precipitation operation (s) , which can be performed in accordance with techniques known per se, e.g. by sulphate precipitation (s) with sulphate ions. The obtained radioactive and/or non- radioactive solids can then be used or disposed of in any suitable manner.

Once radioactive and non-radioactive cationic constituents have been recovered, recovery of dissolving constituents (e.g. chelating agent) can be achieved, e.g. by the appropriate pH adjustment of the effluent solutions to cause precipitation. The filtered chelating agent can be recycled. Residual salt solutions can then be discarded as effluent, though if it is absolutely necessary to have a zero effluent operation, the salt solutions can be reconstituted into acid and alkali by appropriate electrical processes.

According to a second aspect of the invention a simple and efficient apparatus for performing a process as defined above is provided. Said apparatus comprises:

A) a dissolver vessel for a solvent having the ability of dissolving barium and strontium sulphates, together with associated radioactive species, primarily radium, from said scale;

B) at least one anion exchange column to be fed by the solution obtained in said dissolver vessel, for the removal of sulphate ion therefrom;

C) at least one cation exchange column to be fed by the solution from the anion exchange column(s), for ion exchange separation of radium, and optionally other radioactive species, from barium and strontium and optionally other non-radioactive constituent (s) , contained in said sulphate-depleted solution; ; and D) recovery vessels for separated radioactive and non-radioactive fractions from said cation exchange column(s), respectively.

As to preferable embodiments of such an apparatus and the operation thereof reference is primarily made to the preferable embodiments of the process as described above. However, some advantageous apparatus constructions can be summarised as follows: A preferable embodiment of the apparatus is an apparatus wherein said at least one cation exchange column is represented by several cascadingly arranged cation exchanged columns.

Furthermore, the apparatus preferably also comprises a de-oiling unit, arranged before the dissolver vessel, for the removal of the oil content of the starting radioactive oil scale waste.

To accomplish the conversion of the radioactive and/or non-radioactive fraction(s) into solid(s), the recovery vessels are preferably provided with pH regulating means for the precipitation of at least the radioactive fraction into a solid radioactive waste.

The apparatus also preferably comprises a grinding unit, before said dissolver vessel, for reducing solid oil scale to a particulate material, preferably to a particle size of at most 50 μm.

Finally, an apparatus can be referred to, which comprises radioactivity analysis means in connection with the cation exchange column (s) to enable correct collection of fractions.

Drawings

In the drawings the following is shown: Fig. 1 shows a block diagram of one embodiment of the process according to the present invention;

Fig. 2 shows the cascading principle when using several cation exchange columns for the separation stage; Fig. 3 shows a block diagram of one 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;

Fig. 5 shows DCyTA ion exchange as described below in Example 2;

Fig. 6 shows ion exchange separation with EDTA as described below in Example 3 ; and

Fig. 7 shows ion exchange separation with EDTA using a loading and elution technique. More specifically, the block diagram of Fig. 1 shows how oil scale is originally received and subjected to a de-oiling operation. The de-oiled scale is then dissolved as previously described and the solution obtained is, after chemistry adjustment, passed to a separation stage in accordance with the invention, the separated, precipitated "barium" sulphate being obtained in a form for clean disposal and the precipitated "radium" sulphate being volume reduced and in a form for low radioactivity disposal . Furthermore, it is also shown how the solid separated from the dissolution step can be disposed of as an active or non-active residue.

Fig. 2 shows how an input solution is subjected to a first separation stage in a cationic exchange resin, the barium rich and radium rich fractions from said first separation then being passed to separate new, cationic exchange resins, respectively, etc. Barium sulphate and radium sulphate solids are separately recovered for disposal, while radium and barium rich fractions, as shown, are recycled to the first exchange resin.

The block diagram of Fig. 3 shows a heat treatment unit for providing a de-oiled scale material; followed by a dissolver vessel for the dissolution stage; an anion exchange column for chemistry adjustment of the solution obtained in said dissolver vessel; a cation exchange column for the separation stage; and recovery vessels for the radioactive and non-radioactive fractions, respectively.

The operation of such an apparatus should be clear from the process description above and need not be repeated here.

Examples

The present invention will now be described by reference to the following examples :

Example 1 - Dissolution Duplicate oil scale samples from two sources ("A" and "B") were heat treated to pyrolyse organics, and were separately ground to < 50 μm particle size and homogenised using a Retsch (S100) centrifugal ball mill with reversing mechanism. Two 5 g portions of each were accurately weighed and submitted for gamma-spectrometry counting to determine Ra-226 activity. The samples were counted in the same geometry as a 50 Bq standard prepared from 5 g of the natural mineral clinoptilolite spiked with an aqueous certified Ra-226 reference standard. Multiple contact dissolution tests were carried out on 5 g portions of the ground samples. In this technique, the oil scale was contacted with chemical dissolving solution, filtered and then contacted with a fresh batch of solution. The number of contacts was recorded and the cumulative dissolution of activity and weight measured.

The ground oil scale sample was added to 300 cm³ fresh

0,1 M EDTA solution at pH 9 (Na form, pH adjusted with NaOH) . This was maintained at 50 ± 5°C in a 600 cm³ tall glass beaker on a hot plate. The oil scale and chemical dissolving solution were vigorously mixed using a laboratory high shear mixer (Ystral GmBH Type X1020) for

30 minutes. The resulting mixture was centrifuged twice at 4500 rpm for 5 minutes. The supernatant dissolution contact liquor was decanted off and filtered under vacuum through a 0.2 μm glass micro-fibre filter. The filtered dissolution liquors were submitted for 2 hour gamma- spectrometry counting, quantified against an equivalent geometry 100 Bq Ra-226 standard prepared from deionised water spiked with the certified Ra-226 reference standard. The mean loss of non-transferable solid material per filtration of the dissolution contact liquors was 9 mg, determined by accurate weighing of duplicate post-filtration dried filters. The final treated solids were accurately weighed and then made up to 5 g with clinoptilolite in the correct counting geometry. These were submitted for 12 hour gamma- spectrometry counting. The results in Fig. 4 show the extent of dissolution of solid 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 duplicate. In the case of the sample "B", where significant undissolved solid is left, the residual solid is below the radioactivity free release limit. Example 2 - Ion Exchange separation with DCyTA

Column Preparation

A 75 cm³ cation exchange column in the sodium form was prepared from Amberlite IR-120 resin (capacity

1.9 meqcrrf³) . The overall cation capacity of the column was approximately four times the theoretical requirement for the solution barium content, to ensure sufficient available capacity for radium/barium separation in the presence of other metal ions in the dissolution liquor.

Removal of sulphate and pH change in the dissolving solution

A 75 cm³ anion exchange column was prepared using a weak base anion resin, Ionac 365, capacity 2 meqcm^"3.

Assuming up to 4 g barium sulphate potentially dissolved in the dissolution liquor, the overall capacity of the column provided approximately 75% excess. The column was converted into the chloride form from the hydroxide form by passing through of 5 bv (bed volumes) 2 M sodium chloride solution, followed by a 2 bv deionised water rinse.

The first contact dissolution liquor (300 cm³) from an oil scale dissolution of the type described in Example 1 was loaded onto the column at a flow rate of ~1.5 cm³min^_1, followed by a 4 bv deionised water rinse. No precipitation was observed in the column, and the successful removal of sulphate was verified using Merckoquant Sulfat-Test strips, detection limit < 200 mgl^"1. The total column effluent was combined and a

300 cm³ aliquot submitted for counting to ensure that the majority of activity had passed through the resin. The combined solution was spiked with 1000 Bq Ra-226 (from a certified standard in HCl) and again a 300 cm³ aliquot was submitted for counting. The solution was spiked with extra radium in order to facilitate the subsequent cation exchange experiments. One half of the spiked solution (pH 8.5) was reduced to pH 7 for ion exchange separation.

Loading and Separation

The pH 7 adjusted dissolution liquor spiked with additional Ra-226 (described above) was loaded onto a 75 cm³ cation exchange column in the sodium form (pH 7) prepared in the manner described above . After loading of the dissolution liquor the column was rinsed with 2 bv deionised water followed by an additional 1 bv deionised water. 13 bv 0.05 M DCyTA at pH 8.5 was passed through the column to elute the barium taking column effluent fractions for counting and qualitative barium analysis after 5 bv, 10 bv and 13 bv. Subsequently 5 bv 0.05 M EDTA at pH 8.5 was passed through the column to elute the radium followed by a 2 bv deionised water rinse. As previously, each fraction was submitted for counting and qualitative barium analysis in order to track the radioactivity and presence of barium. The majority of the barium was recovered in the first 10 bed volumes of DCyTA effluent. In particular, the first five bed volumes of effluent contained barium but no radium. The radium activity recovery is shown in Figure 5.

Example 3 - Ion Exchange separation with EDTA

The first contact oil scale dissolution liquors from samples A (5) and B(5) were combined and subjected to sulphate removal by anion exchange as described above . The combined anion exchange effluent was adjusted to pH 7 with 2 M HCl. The entire solution (unspiked, combined Ra- 226 activity -500 Bq) was loaded onto a 75 cm³ cation exchange column in the sodium form (pH 7) , prepared in the manner described above. A total of 12 bv (i.e.

900 ml) of solution was loaded. Fractions were taken after 4 bv and every subsequent 2 bv to provide an indication of the point (s) at which radium and barium break-through occurred. Further to loading of the combined dissolution liquors, the column was rinsed with

2 bv deionised water. Subsequently 2 x 5 bv (i.e. total 10 bv) 0.05 M EDTA at pH 8.5 was passed through the column to elute the radium and any residual barium, followed by a 2 bv deionised water rinse. As previously, each fraction was submitted for counting and qualitative barium analysis in order to track the radioactivity and presence of barium. The radioactivity recovered in each fraction is shown in Figure 6. The initial breakthrough of barium from the ion exchange was completely radium- free, and indeed quite large quantities of radium- depleted barium were eluted from the column. The elution stage recovered barium-depleted radium.

Example 4 - Ion Exchange Separation with EDTA using a loading and elution technique

500 ml of a solution of first contact oil scale dissolution liquors was prepared in the same way as in example 3. This sample was also treated by anion exchange (also as in example 3) to remove sulphate ion. Although natural radium was present in this solution extra radium was added in the form of a "spike" to aid subsequent analysis. The barium concentration in the solution was 1.4 g 1-1 and the radium concentration of the solution, after spiking, was 29 Bq ml-1. No adjustment was made to the pH of this solution (which was measured as 5.8) . A cation exchange column (as in example 3) was pre-conditioned by passing through 2 bed volumes of 0.1M EDTA at pH 5. The spiked solution as above was then passed through the cation exchange column.

The solution was measured for radium and barium after passing through the column, and no radium or barium was detected (ie all the barium and radium had been retained on the column. The ion exchange column was then eluted with 10 bed volumes of 0.1 M EDTA adjusted to pH 7, followed by 3 bed volumes of 0.1 M EDTA adjusted to pH

8.5. Fractions corresponding to one bed volume were collected, and measured for barium and radium. The results, shown in Figure 7, indicate a good separation of barium and radium.

Claims

CLAIMS 1. A process for the volume reduction of radioactive oil scale waste, which comprises: a) providing a de-oiled solid scale material; b) subjecting said solid scale material to a solvent having the ability of dissolving barium and strontium sulphates, together with associated radioactive species, primarily radium, therefrom so as to form a scale solution; c) separating any undissolved solid from said scale solution; d) passing the scale solution through an anion exchange resin so as to remove sulphate ion therefrom; e) passing the sulphate-ion depleted solution from the anion exchange resin through a cation exchange resin so as to accomplish separation of radium, and optionally other radioactive species, from barium and strontium, and optionally other non-radioactive constituent (s) , into separate radioactive and non-radioactive fractions, respectively; and optionally f) converting said radioactive and/or non- radioactive fraction(s) into solid form(s) ; the radioactive fraction from the process being volume reduced and in a form suitable for disposal .

2. A process according to claim 1, wherein, in step a) , oil is removed from collected oil scale by means of an operation selected from steam distillation, solvent extraction and heat treatment.

3. A process according to any one of the preceding claims, wherein, in step a) , said solid scale material is provided in particulate form, preferably by grinding, and especially with a particle size of at most 50 μm.

4. A process according to any one of the preceding claims, wherein, in step b) , said solvent is an alkaline solution containing a chelating agent, preferably as the only chemical additive to said solution.

5. A process according to claim 4, wherein said chelating agent is selected from EDTA, DPTA and DCyTA.

6. A process according to any one of claims 4 and 5, wherein the concentration of said chelating agent is within the range of 0.01 - 0.5 M, preferably 0.01 - 0.1 M.

7. A process according to any one of claims 4-6, wherein the pH of said alkaline solution is within the range of 9-11.

8. A process according to any one the preceding claims, wherein, in step c) , said separation is accomplished by settling and filtration.

9. A process according to any one of the preceding claims, wherein said step d) is performed by means of at least one anion exchange column.

10. A process according to any one of the preceding claims, wherein, said anion exchange resin is in the chloride form.

11. A process according to any one of the preceding claims, wherein said step e) is performed by means of at least one cation exchange column.

12. A process according to any one of the preceding claims, wherein said step e) is performed at a pH within the range of 5-7.

13. A process according to any one of the preceding claims, wherein said radioactive and non-radioactive fractions are recovered by passing an elution solution through said exchange resin.

14. A process according to claim 13, wherein said elution solution is based on a chelating agent other than that used for the dissolution operation in step b) .

15. A process according to claim 13, wherein said elution solution is based on the same chelating agent as the one used in the dissolution operation in step b) , the elution being performed at a different pH.

16. A process according to any one of the preceding claims, wherein said step e) is performed by cascading cation exchange columns so as to feed radium-depleted barium and barium-depleted radium to further separate cation exchange columns, respectively.

17. A process according to any one of the preceding claims, wherein, in step f) , said conversion is performed by means of a precipitation operation, preferably by means of sulphate precipitation (s) .

18. A process according to any one of the preceding claims, wherein, after said radioactive and non- radioactive fractions have been separated, dissolving constituent (s) , such as chelating agent (s) , is (are) recovered.

19. A process according to claim 18, wherein said recovery is accomplished by precipitation.

20. An apparatus for the volume reduction of radioactive oil scale waste, which comprises:

C) at least one cation exchange column to be fed by the solution from the anion exchange column (s), for ion exchange separation of radium, and optionally other radioactive species, from barium and strontium and optionally other non-radioactive constituent (s) , contained in said sulphate-depleted solution; ; and D) recovery vessels for separated radioactive and non-radioactive fractions from said cation exchange column(s), respectively.

21. An apparatus according to claim 20, wherein said at least one cation exchange column is represented by several cascadingly arranged cation exchanged columns.

22. A apparatus according to any one of claims 20-

21, which comprises a de-oiling unit, arranged before the dissolver vessel, for the removal of the oil content of the starting radioactive oil scale waste.

23. An apparatus according to any one of claims 20-

22, wherein said recovery vessels are provided with pH regulating means for the precipitation of at least the radioactive fraction into a solid radioactive waste.

24. An apparatus according to any one of claims 20- 23, which comprises a grinding unit, before said dissolver vessel, for reducing solid oil scale to a particulate material, preferably to a particle size of at most 50 μm.

25. An apparatus according to any one of claims 20- 24, which comprises radioactivity analysis means in connection with said cation exchange column (s) .