GB2508010A - Treatment of Radioactive Material - Google Patents

Abstract

The present invention relates to a method for the treatment of a magnesium-containing radioactive material, the method comprises: i) providing a mixture of the radioactive material and water; ii) contacting the mixture with carbon dioxide gas to thereby dissolve at least a portion of magnesium in said radioactive material and form a magnesium-containing solution; iii) passing the solution through an ion exchange unit to recover magnesium ions. Also disclosed is a method for the treatment of a aluminium-containing radioactive material, the method comprises: i) contacting the radioactive material with alkaline solution to dissolve aluminium; ii) separating the dissolved aluminium and alkaline solution from undissolved radioactive material; iii) contacting the aluminium and alkaline solution with carbon dioxide to precipitate aluminium salts.

Description

Treatment of Radioactive Material The present invention relates to a method for the treatment of radioactive waste. In particular, the method allows for the reduction in volume of an intermediate level radioactive waste containing one or both of aluminium and magnesium.

Early nuclear power reactors, particularly those designed for the production of plutonium for nuclear weapons in the period from 1940's to the 1970's, usually relied upon natural uranium metal as the fuel. Any other chemical form of uranium or alloying element would introduce additional neutron capturing components which would increase the difficulty for the reactor designers of achieving a self-sustaining critical assembly. This fuel would be provided within a fuel assembly comprising a core of uranium with a cladding material based on magnesium and/or aluminium. Among the properties of these cladding materials which made them particularly suitable for this duty was their low capture cross section (i.e. transparency) for neutrons. Such fuel assemblies were used in so-called Magnox reactors.

The use of such fuel assemblies was generally found to be satisfactory for operation in these reactors. However, a disadvantage of these types of nuclear fuel assembly was that they were not really suitable for long term storage. The aluminium, magnesium and uranium metals are susceptible to corrosion in water, or could become oxidised in air.

While this disadvantage did not matter too much in early years, because the fuel needed to be reprocessed shortly after withdrawal from the reactor in order to extract the desired plutonium, it is now a considerable concern in the handling of the radioactive waste.

In countries where such fuel assemblies were used (particularly the UK) the fuel cladding was often mechanically removed from the uranium metal fuel as the first stage of fuel reprocessing and the discarded cladding then transferred to a waste store or silo.

In many cases the cladding waste (sometimes called "swarf") is still present in the original waste silos today, many decades after it was initially generated. Significant efforts are now being made to remediate the situation Due to the decades of storage, some of the stored waste has corroded (particularly where it has been stored underwater) to a variable extent and hence is present in the form of a mixture of tangled metallic pieces interspersed in corrosion product sludge.

The original process of mechanically separating the cladding from the uranium fuel was often carried out imperfectly, with the result that a proportion of the uranium metal fuel (with its attendant highly radioactive actinides and fission products) is mixed in with the fuel cladding waste. The uranium is, of course, by now also in a partially or wholly corroded state.

Furthermore, it is common for there to be further contaminants co-mingled with the waste, such as graphite sleeves, thermocouple wires and nose cones together with more general radioactive detritus added to the waste both contemporaneously and subsequently. Nuclear reactors were often graphite moderated thermal reactors and in some cases the fuel assemblies contained graphite components (e.g. struts or sleeves).

Even where the reactors were water cooled and moderated, graphite might be involved in the design (e.g. as a neutron reflector).

There is also waste stored at power station sites, produced when the fuel elements were prepared for transport to the reprocessing plant by a process known as de-splittering" or "de-lugging". In these processes the outer splitter and lug components of the fuel element were removed in order to increase packing efficiency for transport. The resulting waste (principally Magnox alloy, but with other components such as nimonic springs) is known as "Fuel Element Debris" (FED) and needs to be treated as part of the decommissioning process for the relevant power stations. In addition to the storage silos, similar material may be present in fuel cooling ponds originally used for the receipt, storage and transfer of the fuel prior to reprocessing.

Nuclear waste can be classified as low level waste ([LW), intermediate level waste (ILW) or high level waste (HLW). Low-level wastes contain small amounts of mostly short-lived radioactivity. Some high-activity [LW requires shielding during handling and transport but most [LW is suitable for shallow land burial.

Intermediate-level waste (ILW) contains higher amounts of radioactivity and in some cases requires shielding. Intermediate-level wastes include resins, chemical sludge and nuclear fuel cladding components, as well as contaminated materials from reactor decommissioning. It may be encapsulated in cement or polymer materials for disposal.

Short-lived waste may be buried in shallow repositories, while long-lived waste may be deposited in a geological repository.

High-level waste (HLW) is produced by nuclear reactors. It contains fission products and transuranic elements generated in the reactor core. It is highly radioactive and often heat-generating.

Accordingly, it is an object of the present invention to provide an improved method for the treatment of these various forms of waste.

There are various known techniques for the treatment of nuclear wastes, whereby the aim of the treatment is to reduce the volume of the waste by removing low-level waste components. These techniques include chemical dissolution strategies. For example, acidic dissolution has been used to dissolve magnesium from Magnox waste. In such a dissolution process the magnesium constituent of the cladding material is extracted in solution from the waste, which, through the use of filtration, effectively guarantees full and complete separation of the final waste from any residual reactive metal. This results in a reduction in the volume of the radioactive waste and reduces the amount of intermediate level waste (ILW) that needs to be disposed of. However, the technique uses large liquid volumes to fully dissolve the magnesium.

Accordingly, it is desirable to provide an improved method for the treatment of radioactive waste and/or tackle at least some of the problems associated with the prior art or, at least, to provide a commercially useful alternative thereto.

In a first aspect the present disclosure provides a method for the treatment of a magnesium-containing radioactive material, the method comprising: providing a mixture of the radioactive material and water; contacting the mixture with carbon dioxide gas to thereby dissolve at least a portion of magnesium in said radioactive material and form a magnesium-containing solution; and passing the solution through an ion exchange unit to recover magnesium ions.

The present invention will now be further described. In the following passages different aspects of the invention are defined in more detail. Each aspect so defined may be combined with any other aspect or aspects unless clearly indicated to the contrary. In particular, any feature indicated as being preferred or advantageous may be combined with any other feature or features indicated as being preferred or advantageous.

Examples of radioactive materials as used herein in each aspect of the invention include any material having radioactive properties. While the term "waste" is used herein synonymously with the radioactive material, it should be appreciated that any suitable radioactive material can be treated. The methods are especially suitable for the treatment of intermediate level wastes, to reduce the volume thereof and to produce LLW or liquid effluent as a by-product. The most common material will be radioactive waste, such as the by-products of nuclear power generation and other applications of nuclear fission or nuclear technology, such as research and medicine. Examples of the waste that can be treated by the present methods include fuel element debris, fuel cooling pond sludge, Magnox swarf waste and aluminium or magnesium based fuel cladding material.

Examples of suitable wastes include: 1) Primarily Magnesium waste. In this waste stream, which is typical of FED, there is a very high proportion of Magnox alloys (>95%), present as metal or corrosion products. The principal minor constituents of the waste are steel and nimonic alloy items. There is very little irradiated uranium or aluminium present.

2) Mixed aluminium/magnesium waste. In the case where there is greater than 5% aluminium and also greater than 5% magnesium in the waste. Wastes of this nature normally exist at reprocessing plants, such as Sellafield in the UK. These wastes may contain a higher proportion of uranium fuel and its corrosion products (e.g. greater than 1%). These wastes may also contain virtually no graphite (<1%) or significant proportions of graphite (>10%), in cases where the graphite components of the fuel elements formed part of the wastes.

3) Magnesium swarf waste. In this waste stream there is very little aluminium present (c5%) but a higher proportion of steel, uranium fuel and general detritus (e.g. greater than 10%). In this case a solid magnesium product may be the most desirable option.

The first aspect of the present invention relates to a method for the treatment of a magnesium-containing radioactive material. That is the method provides for the reduction in volume of an intermediate or high level radioactive material by the removal of magnesium from the material.

Preferably the radioactive material for the first aspect comprises at least 5wt% magnesium. If the material contains less than Swt% magnesium then the reduction in volume may not merit the processing time. Preferably the waste comprises at least 1 Owt%, preferably at least 5Owt%, more preferably at least 95wt%, even more preferably at least G9wt% or substantially all. That is, certain wastes will contain small amounts of highly radioactive waste mixed with magnesium metal. In these cases the reduction in volume of the waste can be very significant. By magnesium metal, as used herein, it will be appreciated that this also encompasses magnesium-rich alloys.

The method comprises a step of providing a mixture of the radioactive material and water. In some cases a dry material will need to be contacted with, or mixed with, water to provide this mixture. In other cases, such as the treatment of material from cooling ponds or storage silos, the radioactive material may already be in the form of a mixture (or sludge) with water.

The method then includes a step of contacting the mixture with carbon dioxide gas to thereby dissolve at least a portion of magnesium in said radioactive material and form a magnesium-containing solution. The carbon dioxide is preferably sparged (bubbled) through the mixture to ensure even mixing and stirring of the mixture.

The use of carbon dioxide provides a number of advantages. The carbon dioxide is cheap and plentiful, and forms carbonic acid on dissolution in the water and this has a very good selectivity for the dissolution for the magnesium present in the waste.

The method then involves passing the solution through an ion exchange unit to recover magnesium ions. The ion exchange unit allows for the capture and concentration of the magnesium recovered from the radioactive material. Furthermore, carbon dioxide is released in the ion exchange unit and can be recycled to dissolve more magnesium.

Preferably the aqueous effluent from the ion exchange unit is returned to the radioactive material. It will be appreciated that a constant stream of fresh water and carbon dioxide could be used in the present invention. However, the recycling of the water components through the ion exchange unit significantly reduces the volume of the treatment fluids and, hence, the treatment complexity.

Furthermore, recycling the aqueous effluent is especially preferred since it has a number of benefits. For example there may be remaining carbonic acid in the solution, so it is efficient to recycle the solution back to dissolve more magnesium.

Before the step of passing the solution through an ion exchange unit to recover magnesium ions, the solution is preferably filtered. This serves to retain any loose particles or components from the radioactive material. A one or more stage filter is preferably used to ensure that minimal non-dissolved components are carried from the reaction vessel. Such filters and their design are well known in the art.

Preferably the method further comprises a step of regenerating the ion exchange unit with an acid to recover a concentrated magnesium-containing solution therefrom. The use of acids to regenerate ion exchange units is conventional. A suitable acid to perform this regeneration is nitric or hydrochloric acid. In this way it is possible to recover a solution rich in magnesium which can preferably be disposed of.

In conventional dissolution techniques it is allowable for dissolved magnesium ions in solution to be disposed of at sea. In the same way, the solution rich in magnesium ions can be disposed of from the current process, although the volume of the required effluent can be significantly reduced.

Dissolution of magnesium based waste by carbonic acid without the use of an ion exchange unit produces a volume of effluent equivalent to 500 litres for every kilogram of magnesium dissolved. Because of this the rate of processing of waste would be severely restricted. Normally the effluent from dissolution would be processed through pre-existing treatment facilities on the relevant nuclear site which have a pre-determined daily capacity for processing liquid waste. Thus the daily capacity of the effluent facilities therefore effectively limits the rate at which the magnesium based waste can be processed. The present inventors realised that if the magnesium in the waste could be made more concentrated, the processing rate could be increased. The use of an intermediate ion exchange stage as described in the present invention allows a much faster processing rate for magnesium waste, within the constraints of existing effluent treatment facilities.

According to a second aspect of the present invention there is provided a method for the treatment of an aluminium-containing radioactive material, the method comprising: contacting the radioactive material with an alkaline solution to thereby dissolve at least a portion of aluminium in said radioactive material; separating the alkaline solution and dissolved aluminium from the radioactive material; and contacting the alkaline solution and dissolved aluminium with carbon dioxide gas to precipitate one or more aluminium salts.

According to this second aspect the method provides for the treatment of an aluminium-containing radioactive material. That is the method provides for the reduction in volume of an intermediate or high level radioactive material by the removal of aluminium from the material.

Examples of radioactive materials suitable for this method are as discussed above. The method is especially suitable for the treatment of intermediate level wastes, to reduce the volume thereof and to produce LLW as a by-product.

Preferably the radioactive material comprises at least 5wt% aluminium. If the material contains less than Swt% aluminium then the reduction in volume may not merit the processing time. Preferably the waste comprises at least lOwt%, preferably at least 5Owt%, more preferably at least G5wt%, even more preferably at least G9wt% or substantially all. That is, certain wastes will contain small amounts of highly radioactive waste mixed with aluminium metal. In these cases the reduction in volume of the waste can be very significant. It should however be noted that pure aluminium-based wastes are very unusual and wastes containing from 5 to 5Owt% aluminium are more common.

The method comprises a step of contacting the radioactive material with an alkaline solution to thereby dissolve at least a portion of aluminium in said radioactive material. In some cases a dry radioactive material will need to be contacted with, or mixed with, the alkaline solution. In other cases, such as the treatment of material from cooling pools or storage silos, the radioactive material may already be in the form of a mixture (or sludge) with water and the alkaline solution can be added thereto.

The method then comprises a step of separating the alkaline solution and dissolved aluminium from the radioactive material. Preferably the step of separating the alkaline solution and dissolved aluminium from the radioactive material is performed with a filter.

As will be appreciated, it may be possible to decant the aqueous component from the reaction vessel. However, it is preferred that the solution is removed via a filter for procedural efficiency and to minimise the carry-over of any radioactive material.

The method then comprises a step of contacting the alkaline solution and dissolved aluminium with carbon dioxide gas to precipitate one or more aluminium salts. The carbon dioxide is preferably sparged (bubbled) through the solution to ensure even mixing and stirring of the mixture. The aluminium will be precipitated as insoluble aluminium carbonate salts.

The step of contacting the alkaline solution and dissolved aluminium with carbon dioxide gas leaves a supernatant which is preferably returned to the radioactive material. It will be appreciated that a constant stream of fresh alkaline solution and carbon dioxide could be used in the present invention. However, the cycling of the supernatant components significantly reduces the volume of the treatment fluids and, hence, the treatment complexity.

Preferably the supernatant is filtered before being returned to the radioactive material.

This helps to prevent any of the aluminium salts from being carried back to the radioactive material Preferably the alkaline solution comprises an alkali metal hydroxide, preferably sodium hydroxide. While any alkaline solution would be appropriate, in view of the volume of material to be treated and the operational costs, sodium hydroxide represents the most cost efficient source of the alkaline solution. Alternatively, another alkali metal hydroxide could also or alternatively be used.

Preferably the pH of the alkaline solution is from 12 to 15, more preferably from 12.5 to 14. As will be appreciated, the higher the pH, the faster the aluminium will dissolve.

Preferably the step of contacting the radioactive material with an alkaline solution uses the solution at a temperature of from 15 to lOOt, preferably from 25 to 900, more preferably from 50 to 7OC. The greater the temperature used, the faster the aluminium components will dissolve but the greater the processing costs for the elevated temperature. Accordingly, the specific process used will depend on a balance of the processing speed and the costs.

Preferably the method further comprises dewatering and/or drying the one or more precipitated aluminium salts. This reduces the volume of material that forms the LLW that needs to be disposed of. Preferably the method further comprises encapsulating the one or more precipitated aluminium salts in a monolithic form for long term storage or disposal. That is, the aluminium salts may be mixed with cement or the ingredients required to form a cement and filled into containers. When set the LLW is suitable for long term storage in this form.

The method of the second aspect is particularly suitable for processing aluminium nuclear fuel cladding. It has been found that the dissolution of aluminium sludges arising from corrosion of nuclear fuel cladding can be strongly affected by the storage times and conditions of the sludge. In particular forms of aluminium hydroxide initially formed such as gibbsito can become converted over long periods of storage to aluminium oxyhydroxido (Al(O)OH) forms such as boohmito. It has been found that boehmito is far more difficult to dissolve than aluminium tn-hydroxide, gibbsite, and thus it is important to ensure that any dissolution process devised is capable of dissolving aluminium sludge forms which arise as a result of ageing.

In previous work the dissolution of this material in strongly alkaline conditions was severely restricted due to the need to use chemistry suitable to apply in-situ in the existing storage tanks. This led the inventors to extensive studies of the kinetics of boehmite dissolution in order to define a practical dissolution system which could be applied to the waste in reasonable time. However, in applications elsewhere where the waste can be retrieved first and dissolved in purpose-built vessels, the restrictions are lifted and the dissolution can be accomplished successfully as described in the present invention. Advantageously, the present invention provides a dissolution system which can deal with this aged sludge material (boehmite).

According to a third aspect of the present invention, there is provided a method for the treatment of a magnesium-and aluminium-containing radioactive material, the method comprising: contacting the radioactive material with an alkaline solution to thereby dissolve at least a portion of aluminium in said radioactive material; separating the alkaline solution and dissolved aluminium from the radioactive material; adding water to the radioactive material to form a mixture comprising the radioactive material and water; contacting the mixture with carbon dioxide gas to thereby dissolve at least a portion of magnesium in said radioactive material and form a magnesium-containing solution; and separating the magnesium-containing solution from the radioactive material.

As will be appreciated, the method of the third aspect is preferably based upon a combination of the first and second aspects disclosed herein. That is, the present inventors have discovered that a two step process can be used to extract first the aluminium alone from a waste and secondly the magnesium. The two sets of reaction conditions do not interfere with each other and a large reduction in volume of certain wastes can be achieved.

The process is designed to minimise the number of vessels used and, hence, minimise the capital expense of building a processing plant.

Preferably the method further comprises one or both of recovering magnesium from the magnesium-containing solution and recovering one or more precipitated aluminium salts from the alkaline solution and dissolved aluminium, as described above.

Preferably the step of recovering magnesium from the magnesium-containing solution comprises: (i) using an ion exchange unit to recover magnesium ions; or (ii) heating the magnesium-containing solution to precipitate one or more magnesium salts.

Preferably the method further comprises a step of returning to the radioactive material the aqueous solution remaining after recovery of the magnesium and/or returning to the radioactive material the aqueous solution remaining after recovery of the one or more precipitated aluminium salts.

As will be appreciated, the first and second aspects of the present invention may be combined to provide the details of the third aspect. That is, the method of the third aspect preferably comprises an aluminium removal treatment in accordance with the second aspect disclosed herein, followed by a magnesium removal treatment in accordance with the first aspect disclosed herein.

In any of the aspects disclosed herein the radioactive material comprises at least 5wt% magnesium and/or Swt% aluminium by dry weight of the radioactive material. Preferably the waste comprises at least lOwt% of magnesium and aluminium, more preferably at least 4Owt%, more preferably at least 6Owt%, more preferably at least 8Owt%, even more preferably at least 9Owt% or 95wt% of the waste and most preferably substantially all of the material. In general, for a combined treatment, the content of aluminium will be from 5 to SOwt% and the content of magnesium will be from 5 to 95wt%.

The methods disclosed herein rely on the use of the carbon dioxide gas. Preferably the carbon dioxide gas is formed from the combustion of graphite recovered from the treated radioactive material and/or carbon dioxide gas recovered from the ion exchange unit.

A significant proportion of the total waste may be present as graphite. Graphite is actually quite a problematic radioactive waste stream in its own right due to the presence of certain radionuclides such as 3H, 14C and 36C1. If the graphite was irradiated in reactor at a low temperature it may also contain stored Wigner energy, which may present problems in subsequent long term storage or processing. If graphite is separated from the waste it may constitute a further difficult waste stream to manage, on the other hand if it is not separated it may prejudice the management of the rest of the waste.

The radionuclide content of graphite in the silos may actually be quite low due to two factors. First the majority of the materials have now been out of reactor for several decades. Radioactive decay will therefore have significantly reduced the short lived isotope component (e.g. 3H). Second the fuel elements were only in reactor for comparatively brief irradiation dwell times, and hence the production of the problematic isotopes will have been restricted.

If the graphite is separated from the waste and processed to remove Wigner energy this would be an advantage. Graphite can, in principle, be converted to carbon dioxide gas.

Such conversion would guarantee to dissipate the Wigner energy and allow the graphite to be removed from the waste thereby achieving yet further volume reduction and at the same time this provides a useful reagent on site for the methods disclosed herein.

Preferably the methods disclosed herein may further comprise adding one or more getter materials. These can be added to any of the aqueous reagents mentioned in the processes. That is, they may be added to the alkaline solution, the supernatant, the radioactive material and water mix or the aqueous effluent, or into two or more of these.

Getter materials are known in the art for the treatment of liquid mixtures including radioactive materials. The principle of operation of getters" is that materials are added to liquid radioactive waste to selectively absorb particular radioactive elements present in the waste. The chemical quantity of radioactive elements present is usually very small. For example the specific activity of fission product caesium is typically about 1012 Bq g, and because caesium is not a particularly common element in nature, the combined radioactive and non-radioactive caesium may be present in the liquid waste at concentrations of only parts per billion or even parts per trillion. It is therefore possible to add a very small amount of material which selectively converts the caesium in the liquid waste from dissolved ionic to solid form, thereby enabling the caesium to be removed from the waste by filtration. The liquid waste is thereby decontaminated and the resulting solid waste has a very small volume.

Getters may be added to solution in the form of solids or may be formed in the liquid waste by the process of precipitation. An example of the former material is potassium cobalt hexa-cyano ferrate, which is a material well known for its ability to selectively absorb caesium. An example of the latter would be the addition of iron and aluminium to the liquid waste to form an iron and aluminium hydroxide sludge which has excellent properties for the removal of actinides from solution.

Different types of getters are well established in prior art, but can only be used effectively in chemical conditions in which they perform properly. For the purposes of the present invention, only those getters which can perform in the chemical conditions pertaining can be used. Thus, for example, potassium cobalt hexacyano ferrate is not very stable in strongly alkaline solutions, and hence can be used to remove caesium during the magnesium dissolution stage of the present invention, but is not suitable during the aluminium dissolution phase.

Advantageously, the present methods remove the chemical energy from the waste material and to render it more stable for disposal. Magnesium, aluminium and uranium metals are unsuitable materials to store in the long term as constituents of radioactive waste. These metals can corrode in aqueous environments with the production and potential build-up of explosive hydrogen gas. They can burn in air (magnesium can even burn in nitrogen gas) and they can react exothermically with cementitious materials at high temperatures (e.g. in a fire situation) without the need for an external supply of oxygen. Uranium metal can in certain circumstances form unstable uranium hydride which is pyrophoric, aluminium corrodes in alkaline environments (typically used in encapsulation processes for radioactive waste) and causes considerable difficulties in producing a stable encapsulated product. Moreover, since the corrosion products occupy a greater volume than the original metal, the metals corrode in encapsulated packages over long periods of time generating expansive forces, which can compromise the stability of the packages.

Advantageously the present techniques serve to separate radioactive from non-radioactive components in the waste. Only a small fraction of the cladding waste stored in the silos constitutes radioactive material. In particular, aluminium and/or magnesium are usually the major constituents of the waste. These elements, originally chosen for their transparency to neutrons, can be separated from the waste by dissolution and disposed of in non-radioactive condition, or as very low level radioactive waste. The removal of non-radioactive constituents from the waste enables highly significant volume reduction of the waste, which reduces the subsequent costs of treatment and storage of the radioactive waste. A good example of this is the separation of Magnox from nimonic springs. Magnox FED metal waste frequently contains large numbers of those physically small nimonic springs (which were originally used to secure the fuel element in position in the fuel channel in the reactor). The springs contain significant cobalt, and hence contain the isotope 60Co. The Magnox metal is the major bulk of the waste and is essentially non-radioactive. Thus if the Magnox is dissolved in carbonic acid the springs will not dissolve, allowing the small volume of springs to remain as the solid radioactive waste, while the magnesium solution is discarded as non-radioactive waste. Significant volume reduction of the radioactive waste is the result.

Furthermore, the techniques can be used as a means of decontamination. The removal of sludge and active metals can reveal other materials (such as steel) present in the waste. The dissolution chemistry can then be adjusted to enhance the removal of radioactivity from these components. Certain enhancements of the basic dissolution process to adapt it for radionuclide dissolution. Steel and other components in the waste normally fall into two categories, those which formed part of the original fuel elements and which have been irradiated by neutrons, and those which did not. These latter items are especially suitable for decontamination because radioactive contamination exists only on their surfaces and can be removed. After decontamination by the enhanced dissolution process these items can separated out and disposed of as non-radioactive or very low level radioactive waste, further enhancing the volume reduction achievable. Decontamination by enhanced dissolution can also be applied to the silo, vault or fuel pond structure itself (once the bulk of the material has been removed) in preparation for dismantling and demolition.

The handling of large and varied solid "extraneous" materials in the process presents a significant challenge, because they must be introduced into the process equipment and be removed from it. The variable nature of the materials (e.g. long thin "thermocouple" wires, graphite pieces, large blocks of steel) and the radioactive environment in which they must be handled are difficult problems. The nuclear industry has become skilled at handling these kinds of challenges, and in particular experience has been gained in doing this in the context of actual dissolution of real waste on an industrial scale. The methods used may involve the sorting and segregation of specific components to prevent them entering the process in the first place, or the use of remote-handled baskets to collect solid debris from the process vessels The methods of handling the solids are well established in prior art, and can be used as required for the purposes of this invention.

A significant advantage of the techniques of the present invention is the avoidance of strong acid based dissolution conditions. Many of the extraneous materials present in the waste, and indeed the materials of construction of the vaults and silos in which the waste is contained, are susceptible to corrosion and destruction in acid conditions. The pH neutral or alkaline conditions employed avoid this problem, and the use of carbon dioxide provides a selective weak acid condition. In many cases the waste, once it has been taken out of the vault or silo, can be treated but will leave the vault or silo itself with contaminated residues to be dealt with. Because the chemistry of the present invention is compatible with these structures, the process can be used to decontaminate the vault or silo after it has been emptied of bulk waste. The solution resulting from this decontamination can, of course, be processed through the facilities set up to apply the present invention. The same strategy can be used to decontaminate extraneous items removed from the vault or silo.

According to a fourth aspect of the invention there is provided an apparatus for performing the method of the first aspect of the invention, the apparatus comprising a vessel for safely holding a mixture of a radioactive material and water, said vessel comprising means for, in use, bubbling carbon dioxide gas through the contents of the vessel, and an outlet and an inlet in fluid communication with an ion exchange unit, the apparatus further comprising one or more filtration units between said outlet and said ion exchange unit and a pump for circulating fluids between the outlet and the inlet of said vessel.

According to a fifth aspect of the invention there is provided an apparatus for performing the method of the second aspect of the invention, the apparatus comprising a first vessel for safely holding a mixture of a radioactive material and an alkaline solution, said first vessel having an outlet and an inlet, and a second vessel having an outlet, an inlet and means for, in use, bubbling carbon dioxide gas through the contents of the vessel, said outlet of said first vessel in fluid communication with said inlet of said second vessel and said outlet of said second vessel in fluid communication with said inlet of said first vessel, the apparatus further comprising one or more filtration units between said outlet of said first vessel and said inlet of said second vessel and a pump for circulating fluids between the first and second vessels.

According to a sixth aspect of the invention there is provided an apparatus for performing the method of the third aspect of the invention, the apparatus comprising: a first vessel for safely holding a mixture of a radioactive material and an alkaline solution, said first vessel having first and second outlets and first and second inlets, and means for, in use, bubbling carbon dioxide gas through the contents of the vessel, second vessel having an outlet, an inlet and means for, in use, bubbling carbon dioxide gas through the contents of the vessel, and an ion exchange unit having an outlet and an inlet, wherein, said first outlet of said first vessel is in fluid communication with said inlet of said second vessel and said outlet of said second vessel is in fluid communication with said first inlet of said first vessel, said second outlet of said first vessel is in fluid communication with said inlet of said ion exchange unit and said second outlet of said second vessel is in fluid communication with said outlet of said ion exchange unit, the apparatus further comprising one or more filtration units between said first outlet of said first vessel and said inlet of said second vessel and between said second outlet of said first vessel and said inlet of said ion exchange unit, and one or more pumps for circulating fluids between the first and second vessels and the ion exchange unit.

The invention will now be described in relation to the following non-limiting figures, in which: Figure 1 shows a schematic of the apparatus for a first stage of aluminium dissolution.

Figure 2 shows a schematic of the apparatus for a second stage of aluminium dissolution.

Figure 3 shows a schematic of the apparatus for magnesium dissolution.

As shown in figure 1, in the first stage of aluminium dissolution a solution of sodium hydroxide (1) is used to dissolve aluminium from a radioactive waste material (5) in the form of mixed magnesium/aluminium sludge/metal waste from a first vessel (10).

Magnesium does not dissolve at this stage and the supernatant is passed through a filter (15) to the second vessel (20). Filtered material is backwashed to the initial reaction vessel (10).

As shown in figure 2, the sodium aluminate solution passed to the second vessel (20) is sparged with carbon dioxide gas (25) to precipitate aluminium carbonate (40). The liquid supernatant is returned through a second filter (35) to the first vessel (10). The aluminium carbonate sludge (40) is removed as low level radioactive solid waste As shown in figure 3, magnesium is dissolved from a radioactive waste material (5) in the form of mixed magnesium sludge/metal waste from a first vessel (10). When working on a mixed material waste, the liquid supernatant returned through the filter (35) will be present in the vessel.

The solution returned to the first vessel (10) is sparged with carbon dioxide gas (25) to dissolve magnesium. All the residue left is now highly radioactive and can be removed for nitric acid dissolution for reprocessing or solid waste management as intermediate level waste The liquid supernatant is circulated through a third filter (45) to an ion exchange vessel (50) containing cation exchange resin in the hydrogen form. The resin removes magnesium from solution and causes carbon dioxide to be evolved. The carbon dioxide is recycled to the dissolver vessel. Residual off-gas is vented. The liquid from the cation exchange unit is circulated back to the first vessel (10).

The ion exchange resin may be regenerated. The magnesium loaded cation exchange resin is regenerated by passing through a solution of mineral acid (e.g. hydrochloric or nitric acid) followed by rinse water. The regenerated cation exchange resin is now ready for further duty. The regenerant solution is neutralised with sodium hydroxide to a pH suitable to enable discharge. Discharge may be by disposal at sea, for example.

"Getters" may be added to the solution to enhance the removal of radioactivity prior to discharge. If so, an additional filter must be used on the discharge line.

As an alternative to the ion exchange unit (50), there are alternative ways to retrieve magnesium from the solution. The solution returned to the first vessel (35) is sparged with carbon dioxide gas (25) to dissolve magnesium. All the residue left is now highly radioactive and can be removed for nitric acid dissolution for reprocessing or solid waste management as intermediate level waste. The liquid supornatant is circulated through a filter (45) to a vessel which is heated with a heater. The magnesium is precipitated from solution as magnesium carbonate and carbon dioxide is evolved. The carbon dioxide is recycled to the dissolver vessel. Residual off-gas is vented. The solution is cooled and recycled to the dissolver vessel.

The invention will now be described in relation to the following non-limiting examples.

Examples

The first stage of the process involves the dissolution of aluminium components in the waste by introducing the retrieved waste into the dissolution vessel and exposing the waste to a solution of sodium hydroxide. The solution in the vessel is concentrated (Approx 0.1 Molar) and occupies only a small fraction of the vessel volume. Aluminium metal will corrode with the evolution of hydrogen gas, and aluminium hydroxide will dissolve as sodium aluminate.

Appropriate "getter" materials may be added to the solution to minimise the carry-forward of radioactivity to the next stage. The getters may optionally include iron hydroxide (or iron salts precipitated in the alkaline solution) recovered from elsewhere in the process.

The solution then passes through a filter into a second vessel. In this vessel the solution volume is optionally slightly increased by addition of water and the solution is sparged with carbon dioxide gas (optionally generated from gasification of graphite) to precipitate aluminium carbonate solid. The aluminium carbonate solid is removed and retained for any secondary processing as necessary before discarding as low level radioactive waste. The solution is now recycled to the original vessel and the volume increased with water.

There are now two choices to proceed, according to whether the magnesium is required in the form of a liquid effluent or as a low level solid waste. For a liquid effluent the procedure is as follows: 1) The solution is sparged with carbon dioxide gas to dissolve the magnesium component of the waste. Further getter materials are added as required. The resulting magnesium bicarbonate solution is filtered and passed to a cation exchange resin in which the magnesium is removed and the resulting carbon dioxide gas is allowed to evolve. The evolved carbon dioxide gas is recycled to the dissolution stage. The treated solution is recycled to the dissolver vessel. The use of this recycling method gradually dissolves the magnesium (with all the benefits of highly selective "carbonic acid" dissolution) but without creating large quantities of effluent.

After dissolution of magnesium is complete the undissolved solids in the first vessel are the principal residual radioactive waste from the process. It is expected that the volume of this waste might typically be 5% of the original waste volume. The solid waste must be removed from the vessel and can then be encapsulated for storage and burial.

Optionally the solids may be treated before removal by dissolution in a small volume of nitric acid to dissolve uranium and fuel based radionuclides. If this is done the resulting solution can potentially be passed to a nuclear fuel reprocessing plant which will separate the components of uranium, plutonium and high level waste.

If a solid low level form of magnesium waste is required, the alternative procedure is as follows: 2) The solution is sparged with carbon dioxide gas to dissolve the magnesium component of the waste. Further getter materials are added as required. The resulting magnesium bicarbonate solution is filtered and passed to a vessel where the temperature is raised to approximately 90t. The magnesium precipitates as magnesium carbonate and the resulting carbon dioxide gas is allowed to evolve. The treated solution is cooled and recycled to the dissolver vessel.

Production of Boehmite, AlO(OH), for subsequent tests "Filter cake" and powder" were manufactured in the following manner -20.604 g pure aluminium plate was added to 1 litre of 3M hydrochloric acid and stirred until the Al dissolved. The pH was measured to be -0.09 and neutralised using 6M sodium hydroxide. 411 mL NaOH was used to return pH to 7.07. The solution was then filtered and solids collected. The filter cake was made by vacuuming the filtered solids to remove the bulk of the water. The powder was made by vacuuming the filtered solids and drying to a constant weight at 105C. The moisture content of the filter cake was measured on an Ohaus Moisture Analyser which measures sample start mass, dries to a constant mass and then records final mass. Powder X-ray Diffraction Analysis was carried out on a Bruker D8 Advance Powder X-ray Diffractometer operating with monochromated copper radiation. Data were recorded over a continuous two theta range of 5-95° using a 0.036 two theta stop for 4 Tconds per step.

The powder diffraction pattern of the sample clearly indicates that the sample is largely boehmite (AlO(OH)).

Al(O)OH for subsequent experiments was used in the form of filter cake material rather than dried powder. The moisture content of the filter cake was 84%. 5g dry powdered Al(O)OH corresponded to 31g filter cake material.

Boehmite was also made by the same method from an actual nuclear fuel can artef act.

There was no indication that the boehmite so made behaved any differently from that made from pure aluminium plate as above.

Aluminium Dissolution! Re-precipitation. Maanox Dissolution and Re-precipitation 24.42g of filter cake Boehmite, prepared as above (equivalent to about 1.7g Al), 1.9g of Magnox (an artefact from a UK nuclear fuel can containing the alloy Magnox Zr55), 10.lg of carbon steel and 10.lg of graphite was added to a dry 2L glass conical flask equipped with a magnetic stirrer. 2OmL of 6M sodium hydroxide was added to the 2L glass conical flask and the volume made up to 500mL with deionised water.

The mixture was heated to 70°C under constant agitaion for 30 minutes. The initial pH was 13.09. The solution was removed using a pipette and was vacuum filtered at 74°C.

The liquid filtrate was collected and stored as "Aluminium Dissolution 1 A". The pH of the filtrate was 12.45. The filtered solids were dried in an oven at 70°C and weighed (0.051g).

20mL of GM sodium hydroxide was added to the 2L glass conical flask containing the wet carbon steel, graphite, Magnox and residual Boehmite. The volume was made up to 500rnL with deionised water. The mixture was heated to 70°C under constant agitation for 30 minutes. The pH was 13.12. The solution was removed using a pipette and was vacuum filtered at 74°C, the pH was 12.78. The liq.iid filtrate was collected and stored as "Aluminium Dissolution 1 B". The filtered solids were dried in an oven at 70°C and weighed (0.004g). There was no obvious sign of residual undissolved boehmite after the second dissolution.

"Aluminium Dissolution lA' and "Aluminium Dissolution lB' were combined together and carbon dioxide was bubbled through the solution at 2L.min to form a white precipitate.

The white precipitate was filtered and carbon dioxide was bubbled through again. The process was repeated until a white precipitate was no longer formed. The combined white precipitate was dried in an oven at 70°C and weighed The mass of dried precipitate recovered was 4.56g.

lOOrnL of the colourless filtrate was collected and submitted for metal concentration determination by Inductively Coupled Plasma Optical Emission Spectrometry (ICPOES) and Inductively Coupled Plasma Mass Spectrometry (ICPMS). The results (in ppm) are as shown in Table 1, confirming that virtually no aluminium remains in solution following the carbon dioxide bubbling.

Table 1 -Aluminium dissolution filtrate analysis Ca Mg Sr Na K Ni Cr Fe Al Cs Al(OH)3 Filtrate <1 2 <0.01 5450 7 cOOl <0.01 0.19 3.2 0.005 [mgi1] The remaining colourless filtrate was returned to the 2L glass conical flask containing wet carbon steel, graphite and Magnox and was made up to 1 L with deionised water.

Carbon dioxide was bubbled through the solution for a cumulative total of 9 hours.

During and following the addition of carbon dioxide, the Magnox fragments effervesced, implying that the Magnox was dissolving. The solution was left stirring for 5 weeks with periodic injections of carbon dioxide and progressively acquired a slight rusty appearance due to corrosion of the carbon steel. Upon complete dissolution of the Magnox, the remaining graphite and steel were removed and weighed together (20.134g). The solution was then filtered to remove any undissolved solids. The dry residue was weighed. The filtered solution was then heated for 30 minutes at 90°C to precipitate out a white solid. The white solid was removed, dried in an oven at 70°C and weighed (13.323g). The theoretical weight of magnesium carbonate pentahydrate expected would be 13.77g. A sample of solution was submitted for metal concentration determination by Inductively Coupled Plasma Optical Emission Spectrometry (ICPOES) and Inductively Coupled Plasma Mass Spectrometry (ICPMS) and the results are shown

in Table 2.

Table 2-Magnesium dissolution filtrate analysis Ca Mg Sr Na K Ni Cr Fe Al Cs MgCO3 Filtrate <1 230 <0.01 504 7 <0.01 <0.01 0.04 0.06 0.49 [mgi1] 0 Rereat with caesium and strontium Qresent This was a repeat with 0.508g CsNO3 and 0.496g Sr(N03)2 dissolved with the Boehmite.

Magnox fines were used instead of discrete pieces in this example.

The equivalent filtrate analysis as per Table 1 above was as shown in figure 3: Table 3-Aluminium dissolution (with added Cs,Sr) filtrate analysis Ca Mg Sr Na K Ni Cr Fe Al Cs Al(OH)3 Filtrate <1 88 0.48 721 4 <0.01 <0.01 0.64 1.26 353 [nig.i1] 0 The equivalent filtrate analysis as per Table 2 above was as shown in Table 4: Table 4-Magnesium dissolution (with added Cs,Sr) filtrate analysis Ca Mg Sr Na K Ni Cr Fe Al Cs MgCO3Filtrate <1 51 1.18 5950 5 <0.01 <0.01 0.02 0.05 0.18 [mgi1] From these results it can be seen that the strontium did not pass into the filtrate during the aluminium dissolution stage, it was most probably co-precipitated with the aluminium carbonate. The caesium did pass through the aluminium dissolution and reprecipitation stages, but was co-precipitated, either with the carbon steel corrosion product or with the magnesium carbonate. No attempt was made to control the behaviour of caesium or strontium in this example with getters, but despite that both species were removed in solid form prior to generation of the final effluent.

As discussed herein, the methods of the present invention have a number of advantages in the treatment of radioactive materials: Advantageously the methods of the present invention create a process which can dissolve either magnesium or aluminium based waste materials or mixtures of the two.

The process should be able to dissolve metals and corrosion product sludges and separate the aluminium and magnesium from other radioactive materials present.

Advantageously the methods of the present invention can convert the aluminium and magnesium so separated into a non-radioactive or low-level radioactive fraction or fractions which can be disposed of as liquid effluent or solid low-level radioactive waste.

Advantageously the methods of the present invention can process the waste in such a way as to guarantee the conversion of magnesium, aluminium and uranium metals to more stable chemical forms.

Advantageously the methods of the present invention can deal with iron and steel components by a) separation as non-dissolvable pieces and to decontaminate these pieces sufficiently to allow their removal from the waste and disposal as low level or non-radioactive waste to the maximum extent commercially viable or b) if present initially as iron hydroxide sludges to separate the iron hydroxide as non-radioactive or low level solid or c) to use iron hydroxide materials present in such a way that they absorb radioactive materials during processing and they can then form a component of the ultimate radioactive waste after conversion to a minimum volume form.

Advantageously the methods of the present invention can convert graphite to gaseous carbon dioxide, preferably in such a way that the carbon dioxide can then be used as a reactant in the waste dissolution process.

Advantageously the methods of the present invention can convert, in so far as this is possible and commercially viable, the radioactive materials present in the waste, including uranium, into a form (e.g. a nitric acid solution) which would be compatible with feeding to a nuclear fuel reprocessing plant. In this way the radioactive materials would be amalgamated with existing stocks of uranium, plutonium and high level waste.

Advantageously the methods of the present invention can enhance the separation of radioactivity from the aluminium magnesium, steel and graphite components beyond the

extent achievable by prior art processes.

Advantageously the methods of the present invention can reduce the volume of liquid effluent produced by the process to a minimum.

Although preferred embodiments of the invention have been described herein in detail, it will be understood by those skilled in the art that variations may be made thereto without departing from the scope of the invention or of the appended claims.

Claims (25)

1. Claims: 1. A method for the treatment of a magnesium-containing radioactive material, the method comprising: providing a mixture of the radioactive material and water; contacting the mixture with carbon dioxide gas to thereby dissolve at least a portion of magnesium in said radioactive material and form a magnesium-containing solution; and passing the solution through an ion exchange unit to recover magnesium ions.
2. 2. The method of claim 1, wherein an aqueous effluent from the ion exchange unit is returned to the radioactive material.
3. 3. The method of claim 1 or claim 2, wherein, before the step of passing the solution through an ion exchange unit to recover magnesium ions, the solution is filtered.
4. 4. The method of any of the preceding claims, further comprising a step of regenerating the ion exchange unit with an acid to recover a concentrated magnesium-containing solution therefrom.
5. 5. The method of claim 4, wherein at least a portion of the concentrated magnesium-containing solution recovered from the ion exchange unit is disposed of.
6. 6. A method for the treatment of an aluminium-containing radioactive material, the method comprising: contacting the radioactive material with an alkaline solution to thereby dissolve at least a portion of aluminium in said radioactive material; separating the alkaline solution and dissolved aluminium from the radioactive material; and contacting the alkaline solution and dissolved aluminium with carbon dioxide gas to precipitate one or more aluminium salts.
7. 7. The method of claim 6, wherein the step of contacting the alkaline solution and dissolved aluminium with carbon dioxide gas leaves a supernatant, and wherein the method further comprises returning the supernatant to the radioactive material.
8. 8. The method of claim 7, wherein supernatant is filtered before being returned to the radioactive material.
9. 9. The method of any of claims 6 to 8, wherein the alkaline solution comprises an alkali metal hydroxide, preferably sodium hydroxide.
10. 10. The method of any of claims 6 to 9, wherein the step of separating the alkaline solution and dissolved aluminium from the radioactive material is performed with a filter.
11. 11. The method of any of claims 6 to 10, wherein, in the step of contacting the radioactive material with an alkaline solution, the solution is at a temperature of from 15 to lOot, preferably from 25 to 90CC.
12. 12. The method of any of claims 6 to 11, wherein the method further comprises dewatering and/or drying the one or more precipitated aluminium salts.
13. 13. The method of any of claims 6 to 12, wherein the pH of the alkaline solution is from 12 to 15.
14. 14. The method of any of claims 6 to 13, wherein the method further comprises encapsulating the one or more precipitated aluminium salts in a monolithic form for long term storage or disposal.
15. 15. A method for the treatment of a magnesium-and aluminium-containing radioactive material, the method comprising: contacting the radioactive material with an alkaline solution to thereby dissolve at least a portion of aluminium in said radioactive material; separating the alkaline solution and dissolved aluminium from the radioactive material; adding water to the radioactive material to form a mixture comprising the radioactive material and water; contacting the mixture with carbon dioxide gas to thereby dissolve at least a portion of magnesium in said radioactive material and form a magnesium-containing solution; and separating the magnesium-containing solution from the radioactive material.
16. 16. The method of claim 15, wherein the method further comprises one or both of recovering magnesium from the magnesium-containing solution and recovering one or more precipitated aluminium salts from the alkaline solution and dissolved aluminium.
17. 17. The method of claim 16, wherein the step of recovering magnesium from the magnesium-containing solution comprises: (i) using an ion exchange unit to recover magnesium ions; or (ii) heating the magnesium-containing solution to precipitate one or more magnesium salts.
18. 18. The method of claim 16 or claim 17, wherein the method further comprises a step of returning to the radioactive material the aqueous solution remaining after recovery of the magnesium and/or returning to the radioactive material the aqueous solution remaining after recovery of the one or more precipitated aluminium salts.
19. 19. The method of any of the preceding claims, wherein the radioactive material comprises at least Swt% magnesium and/or 5wt% aluminium by dry weight of the radioactive material.
20. 20. The method of any of the preceding claims, wherein the carbon dioxide gas is formed from the combustion of graphite recovered from the treated radioactive material and/or carbon dioxide gas recovered from the ion exchange unit.
21. 21. The method of any of the preceding claims, wherein the radioactive material comprises one or more of: fuel element debris, fuel cooling pond sludge, Magnox swarf waste and magnesium and/or aluminium based fuel cladding material.
22. 22. The method of any of the preceding claims, wherein the method further comprises adding one or more getter materials.
23. 23. An apparatus for performing the method of any of claims 1 to 5, the apparatus comprising a vessel for safely holding a mixture of a radioactive material and water, said vessel comprising means for, in use, bubbling carbon dioxide gas through the contents of the vessel, and an outlet and an inlet in fluid communication with an ion exchange unit, the apparatus further comprising one or more filtration units between said outlet and said ion exchange unit and a pump for circulating fluids between the outlet and the inlet of said vessel.
24. 24. An apparatus for performing the method of any of claims 6 to 14, the apparatus comprising a first vessel for safely holding a mixture of a radioactive material and an alkaline solution, said first vessel having an outlet and an inlet, and a second vessel having an outlet, an inlet and means for, in use, bubbling carbon dioxide gas through the contents of the vessel, said outlet of said first vessel in fluid communication with said inlet of said second vessel and said outlet of said second vessel in fluid communication with said inlet of said first vessel, the apparatus further comprising one or more filtration units between said outlet of said first vessel and said inlet of said second vessel and a pump for circulating fluids between the first and second vessels.
25. 25. An apparatus for performing the method of any of claims 15 to 18, the apparatus comprising: a first vessel for safely holding a mixture of a radioactive material and an alkaline solution, said first vessel having first and second outlets and first and second inlets, and means for, in use, bubbling carbon dioxide gas through the contents of the vessel, second vessel having an outlet, an inlet and means for, in use, bubbling carbon dioxide gas through the contents of the vessel, and an ion exchange unit having an outlet and an inlet, wherein, said first outlet of said first vessel is in fluid communication with said inlet of said second vessel and said outlet of said second vessel is in fluid communication with said first inlet of said first vessel, said second outlet of said first vessel is in fluid communication with said inlet of said ion exchange unit and said second outlet of said second vessel is in fluid communication with said outlet of said ion exchange unit, the apparatus further comprising one or more filtration units between said first outlet of said first vessel and said inlet of said second vessel and between said second outlet of said first vessel and said inlet of said ion exchange unit, and one or more pumps for circulating fluids between the first and second vassals and the ion exchange unit.