WO2007017680A1 - Process for the preparation of a leach-resistant radionuclide containing solid - Google Patents

Abstract

The invention provides a process for adsorption of radionuclides from e.g. liquids contaminated with radioactive waste, and their subsequent conversion into a leach resistant form suitable for long term storage. Bone char is a useful adsorbent in this context.

Description

PROCE F R T LEACH-RESISTANT REDIONUCLIDE CONTAINING SOLID

The present invention relates to a process for the preparation of a leach- resistant radionuclide containing solid, to solids obtainable by such a process, to the preparation of modified adsorbents, and to the use of bone char in such processes.

TECHNICAL BACKGROUND AND PRIORART

Many harmful contaminants exist in the environment, either because they are present naturally (for example, ground water in Bangladesh is known to have very high levels of arsenic), or as the result of the activity of man. An increasing awareness of the effects on the health of human beings and other organisms of such pollutants has given rise to the need to reduce their level in the environment, both by eliminating the sources of pollution and remedial treatment of contaminated areas.

A particular problem is the contamination of groundwater. In many parts of the world, groundwater forms the bulk of the potable water available to the local population. Any contaminants present in the groundwater therefore present a serious risk to health through ingestion. The problem is further exacerbated by the fact that once pollutants have entered the groundwater, they tend to spread therein by various mechanisms including diffusion and tidal flow.

Remedial removal of contaminants from contaminated groundwater is far from easy, as by definition groundwater is present in subterranean aquifers. Approaches fall into two broad categories:

i. Pump-and-treat: groundwater is extracted from below the water table, treated to remove contaminants, and the water returned; ii. In situ methods, such as "trenching", wherein a trench filled with adsorbent is placed within the path of flow of the groundwater, such that water passing through the trench has contaminants removed.

Metals are one type of contaminant associated with groundwater, and the toxicity of heavy metals such as lead, cadmium and mercury is well documented. Of special concern is contamination by uranium and other radioactive materials, as these are harmful in very low concentrations, and persist in the environment for many years. A particular problem associated with waste comprising uranium and other radionuclides is that of storage once removed from the environment. Certain radionuclides remain significantly radioactive for many thousands of years, requiring them to be stored in a very stable form in order to prevent them re-entering the environment. Several long term storage media have been studied.

Glass waste forms can be of a wide variety of compositions,Jncluding silicate glasses, borosilicate glasses, and phosphate glasses (summarised by Lutze, 1988). In principle, radionuclides are evenly dispersed throughout the glass, although noble metal precipitates and, in some cases, crystalline oxide precipitates enriched in radionuclides may be present. Waste loadings (defence and commercial) are typically in the range of 10 to 30 weight percent.

A particular process (called AVM, or Atelier de Vitrification de Marcoule) consists of calcining the solution of fission products and sending the resulting calcinate, at the same time as a glass frit, into a melting furnace. A glass is obtained in a few hours, at a temperature of the order of 1100 ⁰C, and is run into metal containers.

The glass frit is composed mainly of silica and boric oxide together with the other oxides (sodium, aluminium etc.) necessary so that the total formulation of calcinate and frit gives a glass which can be produced by known glassmaking techniques and which satisfies the storage safety conditions (conditions pertaining to leaching, mechanical strength, etc.).

Synroc is a titanium based, polyphase ceramic developed at the Australian National University and the Australian Nuclear Science and Technology Organization. The primary phases are: zirconolite (CaZrTi₂O₇), hollandite-like phase (Ba_{1 2}(AI₁Ti)₈O₁₆), perovskite (CaTiO₃), and titanium oxides (e.g., TiO₂). Minor phases also include titanates and aluminates, as well as noble metals. Processing conditions are sufficiently reducing to form Ru-rich and PdTe-rich phases. The principal components can each accommodate a range of radionuclides. Hollandite can retain fission products such as Cs, Rb, Ba; zirconolite, U, Zr, Np, Pu; perovskite, Sr and transuranics such as Np and Pu.

7/O₂ ceramic matrix waste form was developed at the Kernforschungszentrum Karlsruhe (Adehelm et al., 1988) and encapsulates the waste in a rutile (TiO2) matrix. Waste loadings of simulated waste oxides of up to 12 wt. % were attained in the laboratory scale experiments. The final product has a low leach rate as compared to borosilicate glass, because of the low solubility of TiO₂.

Glass ceramics have been developed at the Hahn-Meitner-lnstitut in Berlin and at the Whiteshell Nuclear Research Establishment in Canada (Hayward, 1988). The Canadian glass ceramic consists of discrete crystals of sphene (the proper mineral name is "titanite"), CaTiSiO₅, within a matrix of aluminosilicate glass.

Monazite developed at Oak Ridge National Laboratory is unique in that it consists of essentially a single phase, monazite (CePO₄) (Boatner and Sales, 1988). This composition can be synthesized for the full range of lanthanide orhophosphates. Typical waste loadings for simulated U.S. defence waste are 20 wt. % (monazite waste form density is in the range of 4.0 to 5.0 gm/cc).

Cementitious waste forms are mainly considered for suitable for low-level waste (Jiang, et al., 1993; Quillin et al., 1994). Although concrete has been used for low-level waste, FUETAP (Formed Under Elevated Temperature and Pressure) concrete was developed at Oak Ridge National Laboratory and the Pennsylvania State University (McDaniel and Delzer, 1988). It is unique in that it uses the inherent heat of the high- level waste (as well as external heat sources) to accelerate curing, to drive off up to 98% of the unbound water and form a hard, dense product of improved (compared to normal concrete) physical properties. Typical waste loadings are on the order of 15 to 25 wt. percent (density is approximately 2 gem^"3). Leach rates are comparable to those of borosilicate glass. In addition to radiation damage of specific phases in cementitious materials, there is the additional possibility of radiolysis of cement-pore water and potential pressurizatiόn of waste canisters due to an accumulation of the radiolytically produced gases. Solid-state radiolysis of hydrated phases may also effect their long-term stability.

Bone char (also known as bone charcoal or bone black) is the product obtained from the calcining of animal bones. Its chemical constituents are principally hydroxyapatite and carbon. Bone char has a very high surface area to weight ratio, and for this reason it is a very good adsorbent of many materials present in aqueous systems. The principle use of bone char is the removal of coloured impurities in the sugar refining process. Bone char has previously been shown to be effective in reducing the level^" of contaminants in water. US4,902,427 discloses a filter cartridge for removing heavy metals from water, comprising a bone char impregnated filter. The cartridges are stated to be useful in the removal of cadmium, lead and mercury from water passed through them.

US6,428,695 discloses an in situ method for the removal of uranium from contaminated groundwater. Trenches are dug through the flow path of groundwater, and permeable reactive barriers are installed. A barrier consisting of bone char and iron oxide is shown to be superior to a barrier consisting of bone char alone in the removal of uranium from contaminated water near an abandoned uranium upgrader.

A particular problem that persists is that previously developed media for the storage of nuclear waste is subject to cracking caused by radiolysis and subsequent leaching of radioactive material.

A further problem that persists is the long term storage of material that has been used to adsorb nuclear waste.

The present invention seeks to overcome / address the problems associated with the prior art.

SUMMARY OF THE INVENTION

According to a first aspect, there is provided a process for the preparation of a leach- resistant radionuclide containing solid comprising steps of: i. bringing into association liquid comprising radionuclide, and an adsorbent to form radionuclide loaded adsorbent; ii. heating said radionuclide loaded adsorbent to give a leach-resistant radionuclide containing solid.

According to a second aspect, there is provided a leach-resistant radionuclide containing solid obtainable by a process according to the invention.

According to a third aspect, there is provided a radioactive waste storage container containing a leach-resistant radionuclide containing solid of the invention. According to a fourth aspect, there is provided the use of bone char in the preparation of a leach-resistant radionuclide containing solid.

According to a fifth aspect, there is provided a process for the preparation of a leach resistant metal containing solid comprising steps of reacting an apatite-containing material, a co-adsorbate and a metal-containing liquid.

According to a sixth aspect, there is provided a leach resistant metal containing solid obtainable by a process of the invention.

According to a seventh aspect, there is provided a process for the preparation of a modified adsorbent comprising reacting an apatite containing material with a metal selected from the group of uranium, rare earth metals, zirconium, and titanium.

According to an eighth aspect, there is provided a modified adsorbent obtainable by a process of the invention.

According to a ninth aspect, there is provided a use of a modified adsorbent of the invention for sequestering a metal.

According to a tenth aspect, there is provided a use of a modified adsorbent of the invention for storage of a metal.

DETAILED DESCRIPTION OF THE INVENTION

Radionuclide

Radionuclide, as used herein, refers to any element with a nucleus that undergoes radioactive decay to emit α, β, or γ radiation. Radionuclides include ²³⁹Pu, ²⁴¹Pu, ²⁴⁰Pu, ²³⁸U, ²³⁵U, ²³⁴U, ²³⁴Th, ²⁴¹Am, ¹⁰⁹Cd, ¹³⁹Ce, ⁶⁰Co, ⁵⁷Co, ¹³⁷Cs, ⁵⁴Mn, ¹¹³Sn, ⁸⁵Sr, ^99mTc, ⁸⁸Y and ⁶⁵Zn.

Preferred radionuclides are uranium isotopes, ²³⁹Pu, ²⁴⁰Pu, ²³⁸U, ²³⁵U and ²³⁴U. Particularly preferred is ²³⁵U. Leach-resistant

The term "leach-resistant radionuclide-containing solid" as used herein refers to a solid material containing at least a radionuclide which does not release a significant amount of radionuclide into the environment over time.

Preferably, the leach-resistant solids of the invention release less than 20 % of the content of radionuclide by weight when placed in an aqueous solution for 22 hours. More preferably, they release less than 10 %. More preferably, they release less than 5 %. More preferably, they release less than 2 %. More preferably, they release less than 1 %. More preferably, they release less than 0.1 %. More preferably, they release less than 0.01 %. More preferably, they release substantially no radionuclide.

Preferably, the leach-resistant solids of the invention release less than 20 % of the content of radionuclide by weight when placed in an acidic aqueous solution (for example 0.01 M HCI) for 22 hours. More preferably, they release less than 10 %. More preferably, they release less than 5 %. More preferably, they release less than 2 %. More preferably, they release less than 1 %. More preferably, they release less than 0.1 %. More preferably, they release less than 0.01 %. More preferably, they release substantially no radionuclide.

Preferably, the leach-resistant solids of the invention release less than 20 % of the content of radionuclide by weight when placed in a basic aqueous solution (for example 0.01 M NaOH) for 22 hours. More preferably, they release less than 10 %. More preferably, they release less than 5 %. More preferably, they release less than 2 %. More preferably, they release less than 1 %. More preferably, they release less than 0.1 %. More preferably, they release Jess than 0.01 %. More preferably, they release substantially no radionuclide.

In the embodiment of the invention wherein the radionuclide is one or more uranium isotopes, preferably the leach-resistant radionuclide-containing solid comprises at least becquerelite, Ca(UOa)₆O₄(OH)₆-SH₂O and/or ianthinite,

[U₂(UOa)₄O₆(OH)₄(H₂O)₄](H₂O)₅.

In an alternative preferred embodiment wherein the radionuclide is one or more uranium isotopes, preferably the leach-resistant radionuclide-containing solid comprises at least phosphuranylite, Ca(UO₂)₃(PO₄)₂(OH)₂-6H₂O Adsorbent

The term "adsorbent" as used herein refers to a material with the ability of removing or sequestering metal (preferably radionuclide) from a liquid and into the adsorbent material.

Preferred adsorbents comprise at least an apatite (a phosphate of calcium). Examples of apatite are fluorapatite, chlorapatite and hydroxylapatite. More preferably, the adsorbent comprises at least hydroxylapatite.

Preferred adsorbents comprise elemental carbon.

More preferred adsorbents comprise an apatite and elemental carbon.

A more preferred adsorbent comprises bone char. A highly preferred adsorbent consists essentially of bone char. A very highly preferred adsorbent comprises bone char having a carbon content of less than 9 % by weight, more preferably less than 6 %, more preferably less than 5 %, more preferably less than 4.5 %. A particularly highly preferred adsorbent comprises bone char having a carbon content of between 3 and 5 % by weight.

In an alternative preferred embodiment, bone char can be employed which is essentially carbon-free. Such a material is prepared by igniting the original bone char at 600⁰C for two days in an air atmosphere to produce a white solid.

Preferably, the adsorbent has a surface area of at least 1 m²g^"1. More preferably, the adsorbent has a surface area of at least 10 m²g^"1. More preferably, the adsorbent has a surface area of at least 100 m²g^"1. More preferably, the adsorbent has a surface area of at least 1000 m²g^"1.

Liquid comprising radionuclide

The term "liquid" as used herein refers to both aqueous and non-aqueous liquids.

The radionuclide as defined above may be present in the liquid as a solution, suspension, emulsion, colloid, slurry or any other known form. It is preferred that the radionuclide is present in the liquid in solution.

Loading the adsorbent

The adsorbent and the metal containing liquid may be brought into association in any way that enables the radionuclide to be adsorbed by the adsorbent.

For example, the metal containing liquid may be passed through a pad, bed or cartridge comprising adsorbent.

Alternatively, adsorbent may be added to metal containing liquid to form a suspension or slurry.

The person skilled in the art will appreciate that it may be desirable or advantageous to stir, mix or otherwise agitate the adsorbent and metal containing liquid so as to promote effective and rapid adsorption.

After the adsorbent has adsorbed at least some metal from solution, it is referred to herein as "loaded adsorbent".

The skilled person will appreciate that it is frequently desirable to reduce the level of radionuclide present in the radionuclide containing liquid (for example in the treatment of radioactive waste). The amount of radionuclide remaining in solution may be monitored by gamma spectrometry, for example using a well-type HPGe detector.

Subsequent to the loading step, the loaded adsorbent may be removed from the liquid. This may be achieved by any suitable means which will be apparent to those skilled in the art. Suitable techniques include filtration, centrifugation, densiometric separation and decanting.

However, it is not necessary for the subsequent step that the loaded adsorbent is removed from the liquid.

Co-adsorbate

In a highly preferred embodiment, the metal is brought into association with the adsorbent in the presence of a co-adsorbate. The co-adsorbate is a material which improves the adsorbent's ability to adsorb radionuclide, and/or reduces the propensity of the metal to leach from the loaded adsorbent.

It is highly preferred that the co-adsorbate is a metal. More preferably, the co- adsorbate is a metal ion. More preferably, the co-adsorbate is a metal ion selected from the group of uranium, rare earth metals, zirconium, and titanium.

Uranium is a very highly preferred co-adsorbate. The skilled person will understand that it may be preferable to employ uranium having reduced radioactivity compared with naturally occurring uranium (i.e. depleted uranium).

Surprisingly, it has been found that when uranium is employed as a co-adsorbate and bone char as an adsorbent, the uptake and leach characteristics of the loaded adsorbent are significantly different from those of the same material prepared in the absence of uranium.

The co-adsorbate may be added to the adsorbent simultaneously with the radionuclide. Alternatively, the adsorbent may be pre-treated with co-adsorbate to give a modified adsorbent. Alternatively, the adsorbent may be loaded with radionuclide and subsequently treated with co-adsorbate.

Preparation of Co-Adsorbate Modified Adsorbent

The processes useful in the preparation of co-adsorbate modified adsorbent are essentially similar to those outlined above for the loading of adsorbent with metal.

In a preferred process, adsorbent (e.g. apatite-containing material) is allowed to react with a metal containing solution. Preferably, the metal containing solution is a metal ion solution, more preferably a solution of uranium ions, more preferably a solution of uranium acetate.

In a highly preferred embodiment, the adsorbent is bone char and the metal is uranium, preferably in the form of uranyl acetate. This modified adsorbent is referred to as uranium modified bone char (UMBC).

The weight ratio of metal to adsorbent is preferably between 0.2 and 1.5, more preferably between 0.25 to 0.5, more preferably between 0.3 to 0.4, most preferably about 0.35.

In one preferred embodiment, the pH of the solution is below 4, preferably below 3, more preferably about 2.

Performing the reaction under the above-mentioned acidic conditions, when bone char is used as the adsorbent and employing uranium as metal co-adsorbate gives rise to a product having quite different characteristics from the product obtained under higher pH conditions, and is referred to herein as H-UMBC (acidic uranium modified bone char).

In further preferred embodiment, the bone-char employed in the reaction is essentially carbon-free.

In a further preferred embodiment, the reaction is conducted in the presence of phosphate, preferably an excess of phosphate. A suitable and preferred source of phosphate is Na₂HPO₄. When bone char is used as the adsorbent, the material thus produced is referred to as P-UMBC.

Heat treatment

The e.g. radionuclide loaded adsorbent is preferably heat treated to transform it into a leach-resistant radionuclide containing solid.

Preferably, the heat treatment is conducted in a liquid, e.g. as a suspension of e.g. radionuclide loaded adsorbent in a liquid. More preferably, the heat treatment is carried out in an aqueous liquid.

In a very highly preferred embodiment, e.g. radionuclide loaded adsorbent is heated treated in an acidic aqueous liquid. Under these conditions, a leach-resistant radionuclide containing solid is prepared quite different from that obtained under neutral or basic aqueous conditions.

In an alternative preferred embodiment, e.g. radionuclide loaded adsorbent is treated with steam ("steam reforming"). Preferably, steam is passed through a radionuclide loaded adsorbent is in a continuous fashion. This has two advantages: volatile contaminants are driven off, and thus separated from the metals or radionuclides, and the metal or radionuclide -loaded adsorbent is converted to leach-resistant radionuclide containing solid simultaneously.

Alternatively, the heat treatment may be conducted in the absence of a liquid.

Preferably, heat treatment is conducted at a temperature higher than 50 ⁰C. More preferably, heat treatment is conducted at a temperature higher than 60 ⁰C More preferably, heat treatment is conducted at a temperature higher than 70 ⁰C. More preferably, heat treatment is conducted at a temperature higher than 80 ⁰C. More preferably, heat treatment is conducted at a temperature higher than 90 ⁰C. More preferably, heat treatment is conducted at a temperature higher than 100 ⁰C. More preferably, heat treatment is conducted at a temperature higher than 110 ⁰C. More preferably, heat treatment is conducted at a temperature higher than 120 ⁰C. More preferably, heat treatment is conducted at a temperature higher than 130 ⁰C. More preferably, heat treatment is conducted at a temperature higher than 140 ⁰C. More preferably, heat treatment is conducted at a temperature higher than 150 ⁰C. More preferably, heat treatment is conducted at a temperature higher than 160 ⁰C. More preferably, heat treatment is conducted at a temperature higher than 170 ⁰C. More preferably, heat treatment is conducted at a temperature higher than 180 ⁰C.

Preferably, heat treatment is conducted at a temperature between 100 and 220 ⁰C. More preferably, heat treatment is conducted at a temperature between 150 and 200 ⁰C. More preferably, heat treatment is conducted at a temperature between 160 and 18O ⁰C.

Preferably, the heat treatment step is conducted at a pressure higher than atmospheric. More preferably, the heat treatment step is conducted at a pressure greater than 100 kPa (14 psi); more preferably, the heat treatment step is conducted at a pressure greater than 700 kPa (101 psi); more preferably, the heat treatment step is conducted at a pressure greater than 1.4 MPa (202 psi); more preferably, the heat treatment step is conducted at a pressure greater than 1.7 mPa (250 psi); more preferably, the heat treatment step is conducted at a pressure of about 2 mPa (300 psi).

Most preferably, the heat treatment is conducted at a temperature of 160 to 180 ⁰C and a pressure of 1.5 to 2.5 mPa. Preferably, the heat treatment is conducted for a length of time to ensure complete conversion of radionuclide loaded adsorbent into leach-resistant radionuclide containing solid. More preferably, the heat treatment step is conducted for at least 1 minute, more preferably at least 10 minutes, more preferably at least 1 hour, more preferably at least one day, more preferably at least one week.

Preferably, the heat treatment step is conducted in an autoclave or other sealed pressure vessel.

Preferably, the radionuclide loaded adsorbent is heated with a microwave energy source.

Further treatment steps

Subsequent to heat treatment, the leach-resistant radionuclide-containing solid of the invention may optionally be subjected to one or more of the following procedures.

Removal of liquid: if the heat treatment step was conducted in a liquid, this may be separated from the leach-resistant radionuclide-containing solid by any means known in the art, for example by filtration, decanting, densiometric separation or centrifugation. Furthermore, it may be desirable to further dry the separated leach- resistant solid, again by conventional means.

Compressing: certain of the leach-resistant radionuclide-containing . solids of the invention having an open, porous structure have a low density (and correspondingly high volume per unit weight) when initially formed by processes of the invention. To minimise the amount of radioactive waste requiring long-term storage, it is preferably to compress the leach-resistant radionuclide-containing solid. Compressing is suitably achieved by the application of external pressure (for example in a hydraulic press).

Roasting: the leach-resistant radionuclide-containing solids may be subjected to high temperatures (e.g. 300 to 1000 ⁰C, preferably, about 600 ⁰C) in the presence of an inert atmosphere or oxygen. This subsequent step converts the initially formed leach- resistant radionuclide-containing solid into a different form.

Encapsulation: the leach-resistant radionuclide-containing solids may be encapsulated in another material, for instance glass (vitrification) or a plastics, material.

BRIEF DESCRIPTION OF THE FIGURES

The present invention will be described in further detail by way of example only with reference to the accompanying figures in which:-

Figure 1 is a graph;

Figure 2 is a graph;

Figure 3 is a photograph of products of the invention;

Figure 4 is a photograph of products of the invention;

Figure 5 is a graph;

Figure 6 is a graph;

Figure 7 is a graph;

Figure 8 is a graph;

Figure 9 is a graph;

Figure 10 is a graph.

The present invention will be described in more detail with reference to the following non-limiting examples.

EXAMPLES

Example 1

Approximately 0.7g of uranyl acetate (UO₂(CH₃COO)₂ ^H₂O - RMM 424.1) was dissolved in 10ml of water and mixed with 1.2g of bone char (sample GAU468/1) in a high pressure microwave digestion vessel. The vessel was sealed and heated in a CEM MDS 2000 microwave oven whilst monitoring the pressure and temperature of the vessel contents. Microwave power was applied until the contents of the vessel reached 200⁰C. At this point the pressure inside the vessel reached 340 PSI. The vessel was maintained at 200⁰C for 60 minutes. At the end of this time, the pressure in the vessel had increased to 400PSI. The vessel was removed from the microwave oven and allowed to cool. The solution was poured into a separate container and retained. The solid was transferred to a filter paper and air dried. Characterisation

Appearance

The product initially appeared off-white in colour. However, on grinding, the material reverted to its initial black colouration suggesting that- the off-white colouration resulted from a coating of reaction products on the surface of the bone char.

X-ray diffraction analysis

X-ray diffraction analysis of the ground solid was performed using a Panalytical X'pert system using Fe-filtered Co k_α line and operating at 35 kV and 35mA. The analysis indicated the presence of the following minerals:

- apatite;

- becquerelite Ca(UOz)₆O₄(OH)₆ SH₂O (RMM = 1970.41 - 72% U by mass);

- graphite;

- a fourth unidentifiable component (< 10% of the sample).

Uranium loading

A known quantity of the ground solid was measured along with a matrix-matched standard by gamma spectrometry. The loading of uranium in the solid was calculated based on the measured ²³⁵U activity.

The U loading of the prepared solid was determined as 24.8% U by weight and represents almost quantitative uptake of the U on the bone char. Assuming that all the U is present as Becquerelite, this would suggest that the mineral represented 34% of the solid by mass.

Leaching studies

Approximately 0.1g of the prepared solid was mixed with 50ml of water. The mixture was allowed to stand with occasional mixing for 4 days. Aliquots of the solution were removed at intervals, filtered through a 0.45μm membrane filter and the U concentration in the filtrate was determined by ICPMS. The leach test was performed at room temperature and at 50⁰C using a thermostatically-controlled water bath. For both leach tests, less that 0.2% of the total uranium present in the solid sample was leached out into solution, even iafter the maximum leach time of 93 hours.

Conclusions

The reaction of uranyl acetate solution with bone char at 200⁰C and pressures of 300 - 400 PSI resulted in the formation of an off-white coating on the surface of the bone- char. X-ray diffraction analysis indicates that this material was most probably Becquerelite. The average uranium content of the solid was determined by gamma spectrometry as 24.8% indicating that Becquerelite comprised 34% of the solid material by mass. The uranium was effectively bound within the solid phase and was not leached out with water both at room temperature and at 50⁰C.

Example 2

Introduction

In Example 1 it was demonstrated that uranyl ions reacted with bone char under elevated temperatures and pressures to produce a non-leachable uranium bearing compound that was subsequently identified as becquerelite. Becquerelite is one of a series of non-leachable uranium compounds that occur naturally and have been shown to be stable over time scales of millions of years. As such they may prove beneficial as a means of immobilisation of uranium-bearing wastes. In addition such uranium compounds may incorporate transuranic and fission product wastes in a stable form and may stabilise radionuclides adsorbed directly to the bone char substrate providing an economic approach to the immobilisation of nuclear wastes prior to long term storage.

This study focussed on three areas

1. Optimisation of conditions for the preparation of stable uranium-bearing compounds from uranyl acetate and bone char with particular consideration of the scale-up of any proposed procedure.

2. Evaluation of the leaching of uranium from the resulting mineral under a range of conditions. 3. Investigation of the retention and leaching of key radionuclides of importance in nuclear waste treatment on the product.

Preparation of uranium-modified bone char

The uranium modified bone char (U-char) was prepared by mixing uranyl acetate, fine bone char powder (40 micron diameter - supplied by Brimac Ltd) and water. The mixture was either heated in a PTFE beaker in a water bath at atmospheric pressure to 90⁰C or heated under pressure in a sealed microwave digestion vessel to the desired temperature / pressure for a specified time. The exact conditions chosen depended on the test and are summarised in Table 1. For sealed vessel experiments, water can be heated to temperatures in excess of its boiling point with the temperature and pressure inside the vessel being related (Figure 1). Although both parameters were monitored the vessel pressure was used to control the reaction conditions in all cases. Figure 1 : shows the correlation between temperature and pressure for the sealed microwave digestion preparation.

The reaction times refer to the total reaction time including the time taken to reach the required temperature pressure. For high temperature experiments, the initial temperature ramping typically required 10 minutes (Figure 2). Figure 2 shows the evolution of temperature and pressure during a typical sealed vessel preparation (note that the pressure does not increase until the temperature reaches 100⁰C).

In tests 8 and 9, a mixture of radionuclides (⁵⁴Mn, ⁵⁷Co, ⁶⁰Co, ⁶⁵Zn, ⁸⁸Y, ¹¹³Sn and ²³⁹Pu) and 0.1ml of IOOOppm Zr, Ce, Fe and Ni were added to the reaction mixture prior to heating. Technetium-99m was also added in test 8. In all tests apart from test

8, the reaction mixture was pH 6. In test 8, acidified radionuclide / stable element tracer solutions were added to the reaction mixture rendering the solution pH 2. In test

9, the pH of the reaction mixture was adjusted to pH 7 through the addition of NaOH prior to heating. Table 1 : Summary of preparation conditions for each test

Mass of

Mass of uranyl bone char acetate pressure time

Test No (g) (g) (PSI) (mins) test i 1.16 0.7512 300 60 test 2 1.2134 0.7345 atm (90⁰C) 60 test 3 1.1472 0.7789 300 20 test 4 1.2086 0.7468 atm (90⁰C) 7 days test 5 1.2527 0.7258 100 60 test 6 1.1281 1.4342 300 60 test 7 abandoned test 8 1.1242 0.7903 300 60 test 9 1.1222 0.7164 200 60 test 10 1.1176 0.7425 30 60 n.m. - not measured

On completion of the test preparation, the mixture was allowed to cool and the solution carefully decanted off the solid and filtered through a Whatman 40 filter paper. The solid was then rinsed with MiIIi-Q water and the washings were filtered and combined. The rinsed solid was dried overnight under an IR-lamp and then weighed to determine the quantity of material produced.

A sub-sample of the solid produced in tests 1 and 9 were ignited for 48 hours at 600⁰C to investigate whether sample ignition affected the chemical composition and leachability of the product.

For test 8, a fine solid was observed suspended in the solution overlying the bone char. This solid was isolated by filtration and retained separate from the bone char solid.

For all tests, the filtered solution was measured by gamma spectrometry to determine the amount of ²³⁵U, ²³⁴Th and spiked radionuclides (where added) remaining in the solution. A well-type HPGe detector was used for all measurements. The resulting spectra were deconvolved and each identified radionuclide quantified using Fitzpeaks software. These results were subtracted from the initial amount of the radionuclides added to the reaction mixture to give the amount incorporated in the solid phase.

X-ray diffractometry

A sample of the solid phase was ground and loaded into a holder. X-ray diffraction analysis of the ground solid was performed using a Panalytical X'pert system using Fe- filtered Co k_α line and operating at 35 kV and 35mA.

Leaching tests

Approximately 0.2g of both the raw and ignited products from test 9 was weighed into a polythene vial and 20ml of leaching solution were added. The mixture was mixed for 22 hours and then filtered through a 0.45μm membrane filter. The ²³⁵U and radionuclide content of the filtrate was then determined by gamma spectrometry. Leaching was performed using 0.01 M HCI, 0.01 M NaOH and water.

Results

Appearance

In all tests, with the exception of test 8, the solid produced via the reaction of uranyl acetate with bone char settled at the bottom of the reaction vessel and was readily separated from the aqueous phase. The product appeared blue-grey in colour but turned black on grinding suggesting that the blue-grey compound was a surface coating (Figure 3). On ignition, the sample turned yellow brown with a loss in mass of 14% (dry / ignited ratio = 1.16). Figure 3 shows the appearance of raw, ground and ignited uranium-modified bone chars.

Increasing the mass of uranyl acetate relative to the bone char in the reaction mixture resulted in a lighter material (Figure 4) although again, the colour reverted to the black of the original bone char on grinding.

In test 8, although the uranium-modified bone char in the bottom of the reaction vessel appeared similar to previous tests, there was an additional fine yellow solid suspended in the aqueous phase. Figure 4 shows a comparison of normal (test 5) and high U loaded (test 6) bone char.

Uranium loading

The efficiency of uranium incorporation into the solid phase is dependent on the pressure/ temperature conditions for the preparation. At 90⁰C and atmospheric pressure the % U incorporated is only 58% but this rises to 96% when the reaction is performed at 170⁰C / 300PSI (Figure 5). The reaction time at 300PSI appears relatively fast with 94% incorporation of U being achieved after only 20 minutes (Figure 6). Figure 5 shows the percentage of U incorporated into the solid phase under various reaction conditions (60 minute reaction times). Figure 6 shows the Proportion of U incorporated at 300PS I against time.

Although U incorporation appears to be inefficient at atmospheric pressures (with only 58% of the uranium in the solid phase after 60 minutes at 90⁰C), the uptake of U increases with time reaching 97% after 7 days. This suggests that the use of the sealed vessel at elevated temperatures simply accelerates the reactions and that it is feasible to conduct the preparations under demanding conditions.

Increasing the quantity of uranium present in the reaction mixture did not significantly affect the uranium uptake. At 300PSI for 60 minutes, 92% of the uranium was incorporated into the solid phase.

For test 8, >99% of the U was incorporated into the solid phases with approximately equal amounts being distributed between the bone char and the fines.

XRD results

X-ray diffraction of the compounds suggested that the main uranium compounds produced under the standard conditions in the solid phase were a mixture of becquerelite, Ca(UO₂)₆O₄(OH-)₆-8H₂O, and ianthinite, [U₂(UOs)₄O₆(OH)₄(H₂O)₄] ^■ (H₂O)₅, on apatite. There appeared to be no significant difference between the compounds produced under atmospheric or elevated pressures. For test 8, where the reaction conditions were more acidic, the uranium compound predominant in the solid phase was phosphuranylite, Ca(UO₂)₃(PO₄)₂(OH)₂.6H₂O. This was either associated with apatite in the bulk solid or in an isolated state in the fines fraction.

Ignition of the standard uranium, modified char (containing becquerelite/ianthinite) resulted in a significant change producing an unidentified compound with a less complex diffraction spectrum.

Radionuclide uptake

Uptake of ²³⁴Th was high in all tests, exceeding 98% irrespective of the U uptake efficiency. Uptake of other radionuclides was assessed as part of tests 8 and 9. For test 8, high uptake efficiencies > 90% were observed for all radionuclides (including ^99mTc) with the exception of ⁵⁴Mn with 84% uptake. The majority of the radionuclides (with the exception of ¹³⁷Cs) appeared to be associated with the bulk modified bone char with < 20% of the total radionuclide inventory being associated with the fine material. Caesium-137 was more evenly distributed between the two phase in a similar manner to that observed for U. Figure 7 shows the distribution of radionuclides between the bone char and fine fractions for test 8.

In test 9, the uptake of radionuclides was again high, with all radionuclides except Cs showing uptake efficiencies > 90%. Caesium uptake was slightly lower, although 83% of the Cs was still associated with the solid phase. Figure 8 shows the uptake of radionuclides on U-modified char obtained in Test 9.

Leaching results

Leaching of radionuclides from the U-modified bone char was performed using 0.01 M HCI, 0.01 M NaOH and water. The initial pH of the three solutions was 2, 7 and 12 respectively. After leaching for 22 hours, the final pH of the solutions had changed to 5, 6 and 12. The decrease in acidity of the 0.01 M HCI leach may have resulted from the reaction between HCI and carbonate known to be present in the bone char.

For all leach solutions, leaching of radionuclides was limited. Co, Cs, Mn and Y were leached to a limited extent (< 10%) in HCI. In water and NaOH the quantity of leached radionuclide was negligible with the exception of Sn and Y (Figure 9). Less than 0.5% of the U initially present in the solid was leached. Figure 9 shows leaching of radionuclides from U-modified bone char using various leach solutions. Following ignition of the sample, the quantity of Co leached appeared to increase slightly under acidic conditions. However, leaching of Sn and Y under alkaline conditions appeared to be reduced and the proportion of U leached remained insignificant (Figure 10). Figure 10 shows the leaching of radionuclides from ignited U- modified hone char using various leach solutions.

Discussion

At atmospheric pressure and 90⁰C the reaction appears to be slow with only 58% incorporation after 60 minutes. By pressuring the reaction vessel and hence increasing the temperature of the reaction over 90% of the U is incorporated into the solid after only 20 minutes and uptake in excess of 95% was observed after 60 minutes. However, even at atmospheric pressure, 97% of the U was incorporated into the solid phase after 7 days. XRD analysis confirmed that the product was similar in all cases. This suggests that the elevated pressures / temperatures achieved in the sealed vessel do not affect the nature of the chemical reaction or the final product but increase the rate of reaction. Although the pressure vessels were used in this study for the majority of preparations in order to reduce preparation times and increase the number of tests that could be performed, it is viable to produce the uranium modified char under less severe conditions enabling commercial preparation costs to be reduced.

The uranium compounds produced during the reaction of uranyl ions and bone char under near neutral conditions appear to be hydrated uranium oxyhydroxides such as ianthinite and, where Ca is present, becquerelite. It-is possible for uranium compounds to convert to new compounds with time and, depending on the final compound, this could result in uranium compounds that are more or less stable. Roasting of the sample following preparation will dehydrate the uranium compound and could potentially stabilise the uranium and associated fission products. Roasting at 600⁰C certainly resulted in the formation of a structurally simpler compound although it was not possible to identify this material.

Under acidic conditions, the phosphate phosphuranylite, Ca(UO₂)₃(Pθ₄)₂(OH)₂.6H₂θ, is formed that partially separates from the bulk bone char to produce a fine solid suspended in the aqueous layer. Under acidic conditions, it is likely that some of the bone apatite is dissolved rendering Ca and phosphate ions available to react with the uranyl ions in solution. The pH of the reaction mixture therefore has a significant impact on the reactions occurring and hence the properties of the resulting uranium compound and requires further investigation.

In all preparations, the uranium compounds appeared to coat the outer surfaces of the bone char. As the reaction is performed immediately after the reactants have been mixed it is possible that there has been insufficient time for the reactants to diffuse into the bone char. Although this does not appear to affect the loading of uranium onto the char, it is possible that at increased uranium loadings, the uranium compounds could separate from the char. Leaving the reactants in contact with the bone char for a period of time prior to heating may result in the uranium compounds being more intimately mixed with the bone char potentially improving the stability of the material.

High uptakes of radionuclides on the uranium-modified bone char were observed for all radionuclides tested and were comparable for the uranium compounds produced under neutral and acidic conditions. Association of all radionuclides (with the exception of Cs) predominantly with the bulk bone char (containing the bone char substrate and 50% of the phosphuranylite) in test 8 may suggest that adsorption of radionuclides to bone char is the dominant uptake mechanism. For Cs, the distribution is more even between the bulk bone char and the fines (consisting of only phosphuranylite) and comparable with the U distribution suggesting Cs substitution into the U compound.

.Leaching tests demonstrated that the incorporated radionuclides were predominantly non-leachable with only small proportion of Co, Cs, Sn and Y being leached under acidic and alkaline conditions. The lack of significant leaching under acidic conditions confirms that the initial removal of radionuclides from solution during the preparation stages is not simply hydrolysis. Leachability of Cs, Sn and Y under alkaline conditions was reduced by roasting the sample although the leachability of Co under acidic conditions- increased slightly. Additional tests are required to determine the long term stability of the uranium compounds and associated radionuclides. Conclusions

The uranium compounds produced from the reaction of uranyl ions with bone char appear to be similar irrespective of the temperatures / pressures used for the preparation although the rate of reaction is significantly affected. The uranium compounds formed under neutral conditions tend to be either hydrated oxyhydroxides of uranium with Ca substitution where Ca is present. Under acidic conditions, leaching of Ca and phosphate from the bone apatite results in the formation of a phosphate, phosphuranylite.

The uptake of radionuclides is high, with Am, Cd, Co, Cs, Mn, Sn, Y and Zn being retained under both neutral and acidic conditions. Preliminary tests demonstrate that Tc is also quantitatively retained in the solid phase under acidic conditions. The resulting materials are stable with minimal leaching of U and associated radionuclides under acidic or alkaline conditions. With the exception of Co, leachability is further improved by roasting the material prior to leaching.

The ease in economic production of these uranium-modified bone chars coupled with the apparent stability and non-Ieachability of the product makes the process an attractive approach for immobilisation of uranium bearing nuclear wastes and could offer an alternative to such processes as vitrification prior to long term repository storage. Further investigations are required to determine the processes underlying the formation of the uranium compounds and changes in the compound composition and stability with time.

Example 3

In this study, the reaction of uranyl ions with bone char under a range of conditions was investigated as well^" as the identification of the uranium products and potential mechanisms for their formation. Also assessed was the incorporation of other radionuclides typical of nuclear wastes into the uranium compounds, the adsorption of these radionuclides directly onto bone char and the stability of the resulting materials. The formation of the uranium compounds via reaction with bone char immobilises many other radionuclides present in the reaction mixture through incorporation within the LJ phosphate. Such a process, coupled with the ability for bone char to adsorb a wide range of contaminants, would provide a cost effective approach for the isolation of both uranium and associated radionuclides from aqueous nuclear waste streams as well as effectively immobilising the radionuclides prior to long term storage / disposal.

Methodology

The bone char (fines) used in all experiments was supplied by Brimac Services Ltd and consisted of the fine black powder with > 90% of the powder < 425 μm. The bone char comprised 70 - 76% hydroxyapatite, 7 - 9% CaCO₃ and 9 - 11% carbon. Carbon-free bone char was prepared by igniting the original bone char at 600⁰C for two days in an air atmosphere to produce a white solid. During the ignition, the material lost approximately 14 wt% through the loss of carbon. Degassed bone char was prepared by ultrasonication of a suspension of 1.1 g bone char in 10ml of water for 30 minutes.

The uranium modified bone char (UMBC_S) was prepared by mixing uranyl acetate solution with fine bone char powder. During the initial characterisation stages, the mass ratio of U : Bone char was varied from 0.2 to 1.5 to determine the optimum U loading on the bone char. Subsequently the U : Bone char mass ratio was fixed at 0.35. The pH of the reaction mixture was typically 4-5. In one test, the reaction was performed under acidic conditions (~pH 2) giving rise to a new product (H-UMBC). Carbon free uranium modified bone char (UMBCi₉) was prepared using the above procedure but replacing bone, char with ignited bone char in the reaction mixture. Preparation of uranium modified bone char was also performed in the presence of excess phosphate to produce P-UMBC. 0.24 g of Na₂HPO₄ (0.0017 moles PO₄) were added to a mixture of bone char and uranyl acetate prior to heating. In element uptake experiments, trace concentrations of stable elements and / or radionuclides were added to the reaction mixture prior to heating with the total concentration not exceeding 0.007mmol / g bone char. The pH of the reaction mixture was adjusted to 4.5 using sodium hydroxide and an acetate buffer. Unless otherwise stated, the mixture was heated under pressure in a sealed high pressure microwave digestion vessel (CEM MDS 2000 oven and UDV vessels) to the desired temperature / pressure for a specified time. In one experiment, the mixture was heated in a PTFE beaker in a water bath to 90⁰C at atmospheric pressure to investigate the rate of reaction at lower temperatures and atmospheric pressures. On completion of the test preparation, the mixture was allowed to cool and the solution carefully decanted off the solid and filtered through a Whatman 40 filter paper. The solid was rinsed and dried overnight under an IR-!amp and then weighed to determine the quantity of material produced. Uranium loadings were determined- by measuring the proportion of uranium that remained in the filtrate using high resolution gamma spectrometric measurement of the ²³⁵U 186 keV gamma line. Identification of the compounds present In the solid phase was performed using X-ray powder diffraction analysis. The solid phase was ground and loaded into a holder. X-ray diffraction analysis of the ground solid was performed using a Panalytical X'pert Pro system using Fe-filtered Co k^ line (λ = 1.79A) and operating at 35 kV and 35mA.

In initial trials, a mixture of gamma emitting radionuclides containing ²⁴¹Am, ¹⁰⁹Cd, ⁵⁷Co, ⁶⁰Co, ¹³⁷Cs, ⁵⁴Mn, ⁸⁸Y and ⁶⁵Zn along with the alpha emitting radionuclide ²³⁹Pu was added to the reaction mixture prior to heating. After heating, the mixture was filtered and the quantity of gamma emitting radionuclides present in the filtrate determined by HPGe gamma spectrometry. The ²³⁹Pu content of the filtrate was then determined following ion exchange purification using alpha spectrometry. In later trials, a mixture of gamma emitting radionuclides and stable elements were included with the reaction mixture. The range of species were chosen to reflect the radionuclides (including activation products, fission products and actinides produced via neutron capture by ²³⁵U or ²³⁸U) typically present in nuclear wastes and which contribute significantly to the radiological impact of the wastes during long term repository disposal. After heating and filtration, the proportion of the species remaining in solution was determined using gamma spectrometry, alpha spectrometry or ICPMS. The labelled UMBC and P-UMBC were leached using 0.1 M HCI, 0.25M NaOH or deionised water. In addition, labelled bone char, UMBQ_g, H-UMBC and ignited UMBC were leached with water. Approximately 0.2g of sample was leached with 20ml of leachate for 4 days at 70⁰C in a water bath. The mixture was then filtered through a 0.45μm membrane filter and the quantity of species leached was determined using either gamma spectrometry or ICPMS.

Results & discussion

Uranium loading capacities, effect of temperature/pressure/time, effect of degassing

In all cases where the reaction was performed in water the solid produced via the reaction of uranyl acetate with bone char settled at the bottom of the reaction vessel and was readily separated from the aqueous phase. The product appeared blue-grey in colour but turned black on grinding, suggesting that the blue-grey compound was a surface coating. SEM imaging confirmed that tha .uranium was present as a surface covering on the bone char. On ignition, the sample turned yellow brown with a loss in mass of 14% (dry / ignited ratio = 1.16). Increasing the mass of uranyl acetate relative to the bone char in the reaction mixture resulted in a lighter material although again, the colour reverted to the black of the original bone char on grinding. The efficiency of uranium incorporation into the solid phase was dependent on the pressure/ temperature conditions for the preparation. At 90⁰C and atmospheric pressure the % U incorporated was only 58% but this increased to 96% when the reaction was performed at 170⁰C. The reaction time at this temperature appeared relatively fast with 94% incorporation of U being achieved after only 20 minutes (Figure 4). Although U incorporation appeared to be inefficient at atmospheric pressures (with only 58% of the uranium incorporated into the solid phase after 60 minutes at 90⁰C), increasing the pressure led to a 97% uptake of U after 7 days. The use of sealed vessel at elevated temperatures clearly accelerates the reactions. Quantitative incorporation of uranium in the solid phase was observed for uranium loadings between 0.2 and 0.9 gU/g bone char (Figure 5). Above this concentration, the uranium loading in the solid phase remained constant giving a maximum U loading of ca 1.2g U/g bone char equivalent to 5 mmol/g bone char. LJ loading capacities were improved by degassing the bone char prior to performing the reaction (Table 4). Loading capacity was significantly reduced to 63% on using ignited bone char (using a 0.35g U /g bone char ratio) suggesting that either the carbon plays an important role in the reaction or that those other components within the bone char had been altered during the heating process. The introduction of additional phosphate resulted in 100% uptake of U at the same U / Bone char loading.

Composition of product

X-ray diffraction of the product showed that the main uranium compound in UMBC produced under the standard conditions was a uranium oxyhydroxide, becquerelite, Ca(UO₂)_BO₄(OH)₆-SH₂O. There appeared to be no significant difference between the compounds produced at atmospheric and elevated pressures. Ignition of the UMBC (containing becquerelite) resulted in the formation of a uranium oxide UO₂₅. When the preparation was performed under acidic conditions, a fine material was formed which remained suspended in the aqueous phase. XRD analysis of the bulk H-UMBC and the fine material gave similar reflectance spectra consistent with phosphuranylite, Ca(UO₂)₃(Pθ₄)₂(OH)₂.6H₂O along with- a low intensity reflection at 10.3A possibly indicating low levels of autunite. The lack of fluorescence under long wave UV light confirmed that autunite was not a significant component of the product. In the presence of phosphate (P-UMBC), the uranium phosphate, chernikovite (UO₂HPO₄)(H₂O)₄, was produced. Fluorescence of the product under long wave UV light was consistent with chernikovite (Van Haverbeke et al, 1996).

Incorporation of elements

Uptake of radionuclides on untreated bone char was generally high with > 95% Am, Cd, Co, Mn, rare earth elements, Sn, Th, Y Zn and Zr being adsorbed by the bone char. Uptake of Tc and Ru were lower at 71% and 79% respectively whilst only 29% of ¹³⁷Cs was adsorbed. Uptake of ⁹⁹Tc and ¹³⁷Cs was even lower on ignited bone char. Incorporation of radionuclides in the UMBC tended to be comparable with adsorption onto bone char alone with the exception of ^99mTc and ¹³⁷Cs. For ¹³⁷Cs the uptake efficiency increased to 88% when bone char was replaced with UMBC. Technetium- 99m uptake appeared to decline slightly to 62%. PIutonium-239 was also efficiently incorporated into the UMBC with > 99% of Pu being removed from solution. Acid- washing the bone char prior to preparing the UMBCR₊ did not affect the uptake efficiency of any elements studied. When the reaction was performed under acidic conditions, the uptake efficiency for both ^99mTc and ¹³⁷Cs increased to > 90% although the uptake of ⁵⁴Mn and ¹⁰⁹Cd fell slightly to ca 80%. When phosphate was added to the reaction mixture, the uptake of all radionuclides was >90%. For UMBCi₉, a reduction in uptake efficiencies for ⁵⁴Mn, ⁶⁰Co and ⁶⁵Zn was observed, with efficiencies declining to 71 , 56 and 85% respectively. However, the lowest uptake efficiency of 7% was observed for ^99mTc. The enhanced uptake of Cs on UMBC (containing becquerelite), H-UMBC (containing phosphuranylite) and P-UMBC (containing chernikovite) compared with bone char alone suggests that incorporation of Cs in the becquerelite, phosphuranylite and chernikovite is the dominant uptake mechanism. The lack of enhanced uptake of Tc as Tc(ViI) on UMBC suggests that incorporation of the pertechnetate anion into becquerelite is not significant and is in agreement with Chen et al (2000) who concluded that substitution of TcO₄ ^" into such compounds would result in underbonding at the U⁶⁺ site and destabilise the crystal structure. However, given this observation, the enhanced uptake of TcO₄ ^" in the presence of chernikovite is unexpected and warrants further investigation.

Leaching of elements and U Leaching of a range of elements from bone char, UMBC (prepared from untreated, acid-washed and ignited bone char), P-UMBC and ignited UMBC using 1% HCI, 1% NaOH and water was investigated. The final pH of the solutions varied depending on the solid being leached and the leaching agent. Leaching of both UMBC and P-UMBC with 0.1 M HCI resulted in a leachate with a final pH of 1.6. Leaching of the same materials with either water or 0.25M NaOH resulted in a leachate pH between 6 and 7. Similar pHs were observed in leachates of UMBC prepared from acid-washed bone char (UMBCH₊) and ignited bone char (UMBC_ig). However, higher pHs of 8 - 9 were observed for water leachates of bone char and ignited UMBC. No significant leaching of elements was observed from bone char or UMBC variants at neutral to alkaline pHs for all elements with the exception of Tc. At pH 8.8, 21% of Tc was leached from ignited UMBC whilst 8% of the Tc was leached from bone char at similar pHs. Leaching of Tc from ignited UMBC suggests that the loss of carbon reduces the ability of the material to retain Tc and is consistent with the proposed role that carbon plays in the initial uptake of Tc. The low leachability of Cs in the presence of Na is unexpected given the proposed uptake mechanism of ion exchange into the interlayers of the uranium minerals. Burns (1999) suggested that such uptake mechanisms would be reversible and would lead to the release of incorporated Cs with time. Such a release is not apparent in this work and may suggest that the Cs is more irreversibly retained in the UMBC. At pH 1.6, significant leaching (40 - 92%) of Cs, Mn, Co, Zn, Cd, Ce, Sm, Eu and Am was observed. Zr was also leached but to a lesser extent, with only 1% and 8% of the Zr being leached from the UMBC and P-UMBC respectively. Leaching of U was lower even under acidic conditions with 5 - 7% being dissolved.

Table 6 : Percentage incorporation of elements into various products

All results are expressed as % of element removed from original solution onto the solid phase pH refers to the final pH of the solution after reaction; n.m. not measured

Table 7 : Leaching of elements from various products under a range of conditions

Conclusions

Reaction of uranyl ions with fine bone char at pH 4-5 and elevated temperatures and pressures results in the formation of a compound analogous to the natural mineral becquerelite. Under more acidic conditions, leaching of phosphate from the hydroxyapatite results in the formation of phosphuranylite whilst the addition of excess phosphate produces chernikovite. The U loading capacity of the bone char was high with up to 0.9gU/g bone char.

Quantitative uptake of elements on both raw and ignited bone char was found for Mn, Co, Zn, Y, Zr, Pd, Cd, rare earths, Th and Am indicating that these elements will adsorb to the hydroxyapatite component of the char. Tc was adsorbed quantitatively to bone char but not to ignited bone char showing that carbon is important in the uptake of this element. Cs was not adsorbed efficiently to either bone char or ignited bone char but was quantitatively retained in the UMBC variants indicating incorporation of this element into the uranium compound. For most elements, there was no significant difference between the uptake observed for UMBC or P-UMBC although for Tc, the uptake efficiency was higher on P-UMBC. It is therefore apparent that a range of mechanisms control the uptake of elements onto UMBC and are responsible for the wide range of species that may be extracted. The results also demonstrate the superiority of UMBC over raw bone char or hydroxyapatite in the non-selective retention of waste radionuclides.

Leachability of U and other elements was low from bone char and UMBC variants at pH 5 - 8. Leaching of Cs, Mn, Co, Zn, Cd, rare earths and Am was observed under more acidic conditions with little difference being observed between the UMBC and P-UMBC implying that the presence of phosphates did not reduce the leachability at this pH. However, even under acidic conditions, the proportion of U leached was low (~7%).

The wide range of species adsorbed / incorporated into the UMBC and the stability of the product indicates that there is considerable potential for the application of this material in the isolation of radionuclides from aqueous nuclear waste streams and the stabilisation of these wastes during storage and disposal. References

Abdel Raouf M.W.& Daifullah A.A.M. (1997). Potential use of bone charcoal in the removal of antimony and europium radioisotopes from radioactive wastes. Ads. Sci. Technol., 15, 559-569.

Burns (1999)

Burns P.C. & Li Y. (2002). The structures of becquerelite and Sr-exchanged becquerelite. Am. Mineral., 87, 550-557.

Chen F., Burns P.C. & Ewing R.C. (2000). Near-field behaviour of ⁹⁹Tc during the oxidative alteration of spent nuclear fuel. J. Nucl. Mat, 278, 225-232.

Choy K.K.H., Ko D.C.K., Cheung C.W., Porter J.F. & McKay G. (2004). Film and intraparticle mass transfer during the adsorption of metal ions onto bone char. J. Colloid Interface ScL, 271, 284-295.

Choy K.K.H. & McKay G. (2005a). Sorption of cadmium, copper and zinc ions onto bone char using Crank diffusion model. Chemosphere, 60, 1141-1150.

Choy K.K.H. & McKay G. (2005b). Sorption of metal ions from aqueous solution using bone char. Environ. Int., 31, 845-854.

Ewing R.C. (2001). The design and evaluation of nuclear-waste forms: clues from mineralogy. Canadian Mineralogist, 39, 697-715.

Van Haverbeke L. V., Vochten R. and Van Springel K. (1996). Solubility and spectrochemical characteristics of synthetic chemikovite and meta-ankoleite. Min. Mag., 60, 759-766.

Wilson J.A., Pulford I. D. & Thomas S. (2003). Sorption of Cu and Zn by bone char. Env. Geochem. Health, 25, 51-56.

All publications mentioned in the above specification are herein incorporated by reference. Various modifications and variations of the described methods and system of the invention will be apparent to those skilled in the art without departing from the scope and spirit of the invention. Although the invention has been described in connection with specific preferred embodiments, it should be understood that the invention as claimed should not be unduly limited to such specific embodiments. Indeed, various modifications of the described modes for carrying out the invention which are obvious to those skilled in chemistry or related fields are intended to be within the scope of the following claims.

Claims

1. A process for the preparation of a leach-resistant radionuclide containing solid comprising steps of: i. bringing into association liquid comprising radionuclide, and an adsorbent to form radionuclide loaded adsorbent; ii. heating said radionuclide loaded adsorbent to give a leach-resistant radionuclide containing solid.

2. A process according to claim 1 , wherein said adsorbent comprises apatite.

3. A process according to claim 1 or 2, wherein said adsorbent comprises elemental carbon.

4. A process according to any preceding claim wherein said adsorbent comprises bone char.

5. A process according to claim 5 wherein said adsorbent consists essentially of bone char.

6. A process according to any preceding claim wherein the radionuclide comprises at least an isotope of uranium.

7. A process according to any preceding claim wherein the heating step ii is conducted at a temperature of at least 100 ⁰C.

8. A process according to any preceding claim wherein the heating step ii is conducted at a pressure of at least 100 kPa.

9. A process according to any preceding claim wherein the heating step ii is conducted for at least 10 minutes.

10. A process according to any preceding claim wherein the heating step ii is conducted in the presence of water.

11. A process according to any preceding claim wherein the heating step ii is conducted in the presence of an acid.

12. A process according to any preceding claim wherein the heating step ii is conducted with microwave heating.

13. A process according to any preceding claim comprising a further step of roasting the leach-resistant radionuclide containing solid.

14. A process according to claim 13 wherein said roasting step is conducted at a temperature of 300 to 1000 ⁰C.

15. A process for the preparation of a leach-resistant radionuclide containing solid, comprising heating a mixture of radioactive material and bone char in steam and optionally oxygen.

16. A leach-resistant radionuclide containing solid obtainable by a process as claimed in any one of claims 1 to 15.

17. A leach-resistant radionuclide containing solid comprising elemental carbon, apatite, bequerelite and/or ianthinite.

18. A leach-resistant radionuclide containing solid comprising elemental carbon, apatite and phosphuranylite.

19. A radioactive waste storage container containing a leach-resistant radionuclide containing solid as claimed in any one of claims 16 to 18.

20. Use of bone char in a process for the preparation of a leach-resistant radionuclide containing solid.

21. A process substantially as described herein with reference to any one of the examples.

22 A use substantially as described herein with reference to any one of the examples.

23. A product substantially as described herein with reference to any one of the examples.

24. A process for the preparation of a leach resistant metal containing solid comprising steps of reacting an apatite-containing, material, a co-adsorbate and a metal-containing liquid.

25. A process according to claim 24 wherein the apatite-containing material is bone char.

26. A process according to claim 24 or 25 wherein the co-adsorbate is a metal selected from the group uranium, rare earth metals, zirconium, and titanium.

27. A process according to any one of claims 24 to 26 wherein the co-adsorbate is uranium.

28. A process according to claim 27 wherein the uranium is in the form of uranyl acetate.

29. A process according to any one of claims 24 to 28 wherein the apatite-containing material is bone char.

30. A process according to any one of claims 24 to 29 wherein the metal is a radionuclide.

31. A process according to any one of claims 24 to 30 wherein the reaction is heated.

32. A process according to any one of claims 24 to 31 wherein the reaction is conducted in the presence of phosphate.

33. A process according to any one of claims 24 to 32 wherein the reaction is conducted at a pH of below 4.

34. A process according to any one of claims 24 to 33 wherein the reaction is conducted at a pH of below 3.

35. A leach resistant metal containing solid obtainable by a process according to any one of claims 24 to 34.

36. A process for the preparation of a modified adsorbent comprising reacting an apatite containing material with a metal selected from the group .of uranium, rare earth metals, zirconium, and titanium.

37. A process according to claim 36 wherein the apatite containing material is bone char.

38. A process according to claim 36 or 37 wherein the metal is in the form of metal ions.

39. A process according to any one of claims 36 to 38 wherein the metal is uranium.

40. A process according to claim 39 wherein the uranium is in the form of uranyl acetate.

41. A process according to any one of claims 36 to 40 wherein the ratio of metal to apatite containing material is between 0.2 and 1.5.

42. A process according to any one of claims 36 to 41 wherein the reaction is heated.

43. A process according to any one of claims 36 to 42 wherein the reaction is conducted in the presence of phosphate.

44. A process according to any one of claims 36 to 43 wherein the reaction is conducted at a pH of below 4.

45. A process according to any one of claims 36 to 44 wherein the reaction is conducted at a pH of below 3.

46. A modified adsorbent obtainable by a process according to any one of claims 36 • to 45.

47. Use of a modified adsorbent as claimed in claim 46 for sequestering a metal.

48. Use of a modified adsorbent as claimed in claim 46 for storage of a metal.

49. Use according to claim 47 or 48 wherein the metal is a radionuclide.