CA1199043A

CA1199043A - Radioactive waste immobilization using ion-exchange materials which form glass-ceramics

Info

Publication number: CA1199043A
Application number: CA000415622A
Authority: CA
Inventors: Robert A. Speranzini; Peter J. Hayward; Robert E. Hollies; Alan Shaddick
Original assignee: Atomic Energy of Canada Ltd AECL
Current assignee: Atomic Energy of Canada Ltd AECL
Priority date: 1982-11-05
Filing date: 1982-11-15
Publication date: 1986-01-07
Also published as: GB8327364D0; GB2133607B; IT1170543B; JPS5999399A; GB2133607A; IT8349227A0; FR2535890A1

Abstract

ABSTRACT A process for producing a stable glass-ceramic product useful for storing radioactive nuclear waste products for extraordinarily long periods of time without danger to the environment, in which liquid wastes having high level radioactivity are passed through an ion exchange medium, such as a titanate to deposit the highly radioactive materials thereon and produce a low level radioactive liquid waste which can be safely disposed of in a less rigorous manner than high level wastes is described. The ion exchange medium carrying the highly radioactive materials is heated, with any additional glass forming materials which may be necessary, to form a melt having a composition in the range: where M is selected from Na, K; M1 is selected from Ca, Ba, Sr; M11 is selected from Al, B, Fe, Cr; M111 is selected from Zr, Sn, Zn and M1V is selected from P, Ta, Nb. This glassy composition is cooled and then reheated to a selected temperature, dependent on the composition, in the range 950°C - 1050°C to effect devitrification into a glass-ceramic product comprising crystalline sphene in a glassy matrix. Preferably most of the radioactive material will report in the crystalline sphene phase.

Description

CKGROUND OF INVENTION
This invention relates to a process Eor the permanent environmentally and biologically safe storage, for extra-ordinarily long periods of time, of highly radioactive waste materials. More specifically this invention relates to a process for preparinq a glass-ceramic composite product for containment of solid radioactive wastes and to the glass-ceramic product.
Considerable effort in recent years has been directed at developing techniques for conditioning of liquid radio-active wastes for disposal. In particular high-level and intermediate-level liquid wastes from fuel reprocessing are contaminated with fission products and long-lived lanthanides and actinides and must be solidified for disposal. The solidified waste products must be stable enough and disposed of in such a way as to prevent the biologically hazardous radionuclides from contaminating the environment over their hazardous lifetimes. One method which has been considered for disposing of liquid radioactive wastes involves incorporating the wastes directly in glasses, glass-ceramics or ceramics, then emplacing the waste products deep under-ground in geological formations of salt, granite, shale or basalt. In this case inactive constituents of the wastes, which can sometimes adversely affect product durabilities, are incorporated in the products along with the radionuclides so that relatively large volum~s of waste must be disposed of deep underground. Another method which has been considered involves removing long-lived fission products, lanthanides 1~99~9L3 and actinides from the liquid wastes using ion-exchange materials, then converting -the contaminated ion-exchange materials into glasses or ceramics. In this case only relatively small volumes of ion-exchange materials contamin-ated with long-lived con-taminants need be converted into stable, leach resistant products for disposal deep under-ground. The large volumes of decontaminaied liquids, now containing only relatively short-lived radionuclides, could be disposed oE in a less rigorous way as low-level 10 wastes.
Of the ion-exchange materials which have been identified for use in decontaminating radioactive liquid wastes, zeolites, titanates and calcium hydroxyapatite have been shown to be effective at removing cesium, strontium, some lanthanides and some actinides. Attempts have been made to convert the zeolites and titanates to stable leach resistant products suitable for disposal of radioactive wastes. Such attemp-ts have included melting the zeolite with a borosilicate glass, hot pressing calcium titanate to form a ceramic product, and cold pressing mixtures of radioactive contaminated zeolites (Na-form mordenite) and sodium/ammonium titanates followed by atmospheric sintering of the mixtures.
~1though the technology of glassmaking is well established on an industrial scale and the procedures for incorporating ion-exchange materials in glass are straight-forward, glass products are not thermodynamically stable and may be expected to dissolve slowly under the conditions likely to be encoun-tered over very long periods of time in a waste material repository. While great care is taken with the selection of -the repository --likely a granite pluton such as may be found in the Canadian Shield-- there can be no guarantee that ground waters will not permeate the repository and, at the temperatures generated within the stored waste radioactive material r leachingand/or structural instabilities are to be expected.
While certain phases of the ceramic products are generally thermodynamically stable under the expected conditions and are therefore suitable for the storage of radioactive wastes, most ceramic products are mixtures of many crystalline phases only some of which can be identified.
In these cases it is difficult to use thermodynamics and geological data to assess long term stability and compat-abllity oE ceramic products with host rock formations.
Further, the -technology of hot isostatic pressing is complex and is not well developed on an industrial scale. The products made by cold pressing and sintering tend to be porous and have poor mechanical strength. For sintering to produce an impermeable body, long sintering times and high temperatures are usually necessary and the loss of volatile nuclides is correspondingly high. Fabrication of a large ceramic body is notoriously difficult, mainly because the thermal gradients in a ]arge monolith during heating give rise to non-uniform phase formation, differential shrinkages etc. It is doubtful whether ceramic monoliths of good qua]ity and of a size comparable with a standard waste canister for a vitrified product could be made.
S MMA Y OF THE_ NVENTION
In view of all of these limitations there is a need ~9~

for alternative technology for the very long term s-torage of radioactive wastes, and i-t is an object of the present invention to provide a process for the production of a glass-ceramic composite product in which the crystalline S phase is thermodynamically stable and is compatib]e with the host roc~. The glass~ceramic product has sufficient mechanical strength an~ is relatively easy to make in the required size.
Thus, by one aspect of the present invention there is provided a process for preparing a glass-ceramic product for storing radioactive nuclear waste materials for extra-ordinarily long periods of time comprising:
(a) passing liquid radioactive waste materials through a selected ion exchange medium to thereby deposit said radioactive materials thereon and separate a relatively low level radioactive liquid for subsequent disposal;
(b) heating said ion exchange medium bearing said radioactive materials with sufficient glass forming constituents being present, so as to form a meit, the non-radioactive portion of which has a composition in the range M2OO - 15 wt. %
M112O30 - 15 wt. %
SiO235 - 65 wt. %
Tio210 - 35 wt. %
MlO0 - 15 wt. %

CaO5 - 10 wt. %
Mlllo - 3 wt. %

Ml 250 - 3 wt. %

where M is selected from Na, K
Ml is selected from Ca, Ba~ ';r Mll i5 selected from Al, 8, Fe, Cr Mlll i9 selected from Zr, Sn, Zn and MIV is selected from P, Ta, Nb;
(c) cooling saicl melt so as to form a glass; and (d) heat treating said glass so as to crystallize sphene crystallites in a protect;ve glassy matrix and containing said radioactive materi~als.
By another aspect there is provided a glass-ceramic product for storing radioactive nuclear waste materialR for extraordinarily long periods of time, having a composition:
M2O 0 - lS wt.%
M112o3 0 - 15 wt,%
15 SiO2 35 ~ 65 wt.%
TiO2 10 - 35 wt.%
~lo 0 - 15`wt.

CaO 5 - 10 wt.%
Mlllo - 3 wt.%

20 MlV2O5 ~ 3 wt.%
where M is selected from Na, K; M is selected from Ca, Ba, Sr;
M is selected from Al, B, Fe, Cr; Mlll is selected from Zr, Sn, æn; and MlV is selected from P, Ta, Nb; and compri~ing sphene crystallites in a protective qlassy matrix.

By yet another aspect there is provided a cartridge for the treatment of liquid radioactive nuclear waste materials containing an ion exchange medium which, upon heating with sufficient glass forming ingredients forms a glassy product having a composition in the range ,, ~

.?4~,13 - 5a -M2O0 - 15 wt.
M112O30 - 15 wt.
SiO235 - 65 wt.%
TiO210 - 35 wt.
M O0 - 15 wt.
CaO5 - 10 wt.~
Mlllo - 3 wt.%
MlV2O5o - 3 wt.%
where M is selected from Na, K; Ml is selected from Ca, Ba, Sr;
Mll is selected from Al, s, Fe, Cr; Mlll is sel~cted from Zr, Sn, Zn; and ~1 is selected ~rom P, Ta. Nb.
BRIEF DESCRIPTION OF THE DRAWINGS
The invention will be described in more detail with reference to the drawings in which:-Figure 1 is a sketch of the phase diagram for the 3-phase system SiO2, - CaO - TiO~; and Figure 2 is a sketch of a flow diagram of one embodiment of the present process.
DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS
Thermodynamic calculations and petrological-geochemical observations have indicated that sphene, CaTiSiO5, is an ideal host for most nuclear waste ions.
Sphene is a common accessory mineral in granites and granodiorites, both of which are candidate rock types for a nuclear waste repository vault, and is resistant to chemical alteration. Analyses of mineral specimens also indicates that the CaTiSiOS structure is capable of accepting a wide range of solid solution substitutions such as:
For Ca : Th, U, rare earthelements,Na, ~n, Sr, Br For Ti : Al, Fe~+, Fe++, Mg, Nb, Ta, V, Cr, Sn, Zr, Zn, Pb . ~

For O : O}-l, F, Cl.
Studies have also shown that sphene should be stable in the environment of a vault in areas such as the Canadian Shield and which is flooded with typical ground waters at a depth of 1 km or more. Thus, a sphene-based waste form is believed to be an excellent host for wastes resulting from such opera-tions as CANDU fuel recycling under the currently proposed disposal conclitions.
A glass-ceramic composite produc-t has been selected as the best compromise between ~he desirable properties of crystalline materials and the more forgiving nature of glass with respect to compositional variation and radiation damage and the fact that it may be more easily prepared, via glass making techniques, than a purely ceramic product. ~n a preferred embodiment of the present invention, the glass-ceramic comprises CaTiSiO5 crystallites containing at least a substantial proportion of the longer-lived waste ions in solid solution, surrounded by a durable glass matrix.
The CaO-TiO2-SiO5 phase diagram shows the primary sphene crystallization field to be almost adjacent to a large li~uid immiscibility area, and to be surrounded by relatively low-temperature boundary lines in the range 1300-1375C. The area of glass formation in this system is shown in Figure 1. However, the glasses within this area are potentially unstable, and either crystallize or exhibit subliquidus immiscibility during cooling. In the case of glasses that undergo phase separation during cooling, rapid crystallization of CaTiSiO5 occurs during reheating. This crystallization occurs uniformly throughout the bulk of the glass (as opposed to com~encing at the surface), indicating that the phase separation can serve to nucleate the growth of sphene crys-tals.
Additions of ~a20 and Al203 to CaO-TiO2-SiO2 glass formulations have been showrl -to strongly influence the rate and extent of phase separation in the qlass. Additions of Na20 reduce the tendency to separate, whereas A1203 additions promote the separation. One possible explanation for this is as follows.
The Ti ions have been shown by electron spectro-scopic analysis to exist within the ylass network predomin-antly as octahedrally coordinated (TiO6) groups and these are stabilized by the close proximity of charge-balancing Na+ or Ca2+ ions. ~lowever, the addition of A13 ions, which enter the network as (A104) groups, creates a rival demand for the charge-balancing cations, resulting in destabilization of the network and an increased tendency for phase se~aration.
Auger electron microanalysis of the separated phases in a typical Na~O-CaO-A1203-TiO2-SiO2 glass has confirmed the existence of an Al-rich phase and a Ti~rich phase.
Thus, by varying the Na20:A1203 ratio in these glasses, the degree of phase separation can be controlled.
This can, in turn, influence the rate of sphene crystallization, the possible formation of additional minor crYstalline phases, and the glass-ceramic microstructure. The Na20:A1203 ratio also has a significant influence on melt viscosities, although typical melts are extremely fluid, as a result of the high ~rio2 contents.

Crystallir~.ation of a representative parent ~1ass, ,t mole % composition 6.6 Na2O, 5.1 A12O3, 16.5 CaO, 14.8 TiO2, 57.0 SiO2, to give a sphene-based glass-ceramic, is achieved by reheating the ~lass at a temperature between about 950C
and the melting point of the glass, and preferably in the range 950-1050C for 0.50-1 hour. X-ray diEfraction phase analysis and energy dispersive X-ray spectroscopic elemental analysis have confirmed that sphene is the only crystalline phase produced in the composition ranges under s-tudy, and that the matrix consists of residual Na~O-CaO-A12O3-SiO2 glass. Presumahly, the extreme sluggishness of structural reorganization in high-alumina glasses, as evidenced by their high viscosi-ties, is responsible for the absence of further crystallization of, for example, aluminosilicate phases.
]5 As noted above sphene may be prepared synthetically by heating mixtures of the appropriate chemicals in a crucible at about 1400 C for ~bout 1 hour, cooling the glass thus produced and reheating in the range 950C - 1050C to effect devitrification. The precise crystallization temper-ature selected is of course dependent upon the precise composition of the glass selected. The source of the chemicals is largely immaterial and it has been found that certain inorganic ion exchange media, such as zeolites, sodium titanate and calcium hydroxyapatite, which are useful for ion exchange processes on high level radioactive liquid waste materials, can be heated to between about 1250C and about 1600C and preferably at about 1400C to form a melt which has a composition, which can be adjusted by the addition of conventional glass making constituents as necessary, suitable for the production of sphene therefrom.
Typically zeo]ites such as mordenite (~a-form), Zeolon~ 900 (synthetic mordenite Na8A18Si40O96^24 El2O, supplied by Norton Company, Akron, Ohio) are useful for this purpose.
Thus, a cartridge containing a zeolite, such as Zeolon~ 900 and/or soaium titanate and which contains an insoluble source of calcium, such as wollastonite may be used to effect an ion exchange reaction on a high-level radioactive waste material. The high level radioactive materials are absorbed onto the zeolite and titanate and a relatively low level radioactivity liquid waste is discharged for disposal in a less rigorous manner. The entire cartridge is then heated to about 1400C to produce the desired melt, cooled and then reheated to the range 950-1050C, preferably about 1000C so as to effect devitrification. As the entire ion exchange canister or cartridge is used to e~fect the decontamination and then is treated to produce first the glass and secondly the glass-ceramic product, the handling of radioactive materials is minimized and simplifiedO It will be appreciated that other sources of calcium can be used.
For example calcium hydroxide is soluble in aqueous solutions, and could interfere witn the ion exchange reaction: it therefore rnay be added as a solution, after the decontamina-tion reaction is completed, and before heating to form the melt.
One way to supply Na, Ca, as well as Si would be to use a commercially available glass frit; e.g. a typical soda-lime silica frit would be 14 wt.~ Na2O, 9 wt.~ CaO, - 9a -73 wt.~ SiO2 and A wt.% other oxides. An advantage of this approach is that the g]ass ceramic would Melt at lower temperatures: i.e. the glass frit would melt at a relatively low temperature (1250C-1350C) and the other ingredients would dissolve in the melt.
Example 1 Some tests were undertaken to determine the preferred range of compositions for sphene-based glass-ceramics.
Products with compositions listed in ~ables 1 and 2 were prepared by heating mixtures of p~wdered chemicals in platinum crucibles at 1400C for 1 hour.

Table 1 Some Compositions for Sphene-Based Glass Ceramics _ _ _ . ... . .. . . _ Composition _ _Comp nent (wt._~
# Na2O CaO Ti2A123 Si2 P2o5 - .-. - --.-- -__ ____ _ ___ A 8.33 10 66 10.67 5.61 63.94 0.78 B 9.17 11.11 18.01 4.56 56.33 0.81 C 7.91 15.77 13.50 4.~5 57.59 0.78 D 8.20 13.79 14.38 4.71 58.].3 0.79 E 9.02 14.00 18.74 4.20 53.25 0.80 F 9.52 14.14 22.16 3.82 49.56 0.80 .. .. . _ ... . . _ _ . .
Table 2 The Compositions Eor the Sphene-Based Glass-Ceramics of Table 1 Expressed as Wt.% Mordenite (Na-form), Sodium Titanate, Bone Char and Wollastonite _ _ _ _ _ Composition _____ Component (Wt.~) _ # Mordenite Sodium Bone Wollastonite (Na-form) l'itanate Char ~ ___ _ A 65.3 16.3 4 14.4 B 50.9 26.4 4 18.7 C 51.~ 20.6 4 23.5 D 54.4 21.8 4 19.8 E 48.0 28.1 4 19.9 F 43.2 32.9 4 19.9 _ __ _~ _ ___ __ The melts were coole~ rapidly to 750C by casting on an iron plate, then r~heate~ to 1000C at the rate of 5C per minute.
The temperature was held at 1000C for 1 hour, then -the furnace was turned of f to allow the glass-ceramics to cool.

All of the mix~s mel.e(l easily at 1400C, alth~i~yh ~ompositi.c~i A was very viscous. 'rhe sharpest crystallization peaks as determined by Differential Thermal Analysis were recorded for compositi.ons B, E and F at temperatures of 940 C, 900 C
and 960 C respectively. The sole product of crystallization as determined by X-ray Diffraction was sphene in all compositions. ~n -the basis of the results shown in Tables 1 and 2 the range of compositions for sphene-based glass-ceramics prepared with mordenite (Na-form) sodium titanate, calcium hydroxyapatite and wollastonite were estimated to be:

Component Wt.%
Mordenite (Na-form) 48% - 20%
Sodium titanate28% - 12%
Calcium hydroxyapatite 4% - 4%

Wollastonite 20% - 10%.
The following examples illustrate the production of sphene using ion-exchange materials.
Example 2 A mixture comprising

2.2 g sodium carbonate 1.9 g alumina 12.7 g silica

3.4 g titanium dioxide 3.4 g calcium oxide 1.3 g potassium titanate was heated in a platinum crucible at 1400C for 1 hour, cooled rapidly to 750 C by casting on an iron plate, then reheated to 1000 C at the rate of 5 C per minute. The temperaT~re was h~ld at L000C for 3 ~OUI-S, then Ihe furnace was switched off to allow the glass ceramic product to cool slowly. The product had a composition:
Component T~t.i-Na~O 5,4 K2O 1.2 A12O3 7.9 ~io~ 53.1 TiO2 18.2 C~ 14.2.

It was deterlllined by X-ray diffraction analysis that the only crystallization product was sphene.
Example 3 15 A powdered mixture comprising:
55.0 q Zeolon(~) 900 (sodium form) 22.0 g Sodium titanate 20.0 g ~ollastonite (CaSiO3) was heated in a ceramic mould at 1400C for 2 hours, cooled rapidly to 750C by removing the crucible from the furnace, then reheated to 1000C at 5C per minute, soaked at 1000C
for 3 hours and furnace cooled. The glass-ceramic product had a composition:
Component ~t '~
Na2O 8.3 SiO2 ~5.8 rrio2 18.3 CaO 9.9 and X-ray diffractioIl analysis showed sphene as the sc,i~
crystalline produc-t.
Example 4 . . .
A powdered mix-ture comprising 65.0 g Zeolon('~! 900 (Na-form) 23.0 ~ Sodium titanate 17.2 g Calcium hydroxide was heated in a ceramic crucible at 1400 C for 1 hour, cooled rapidly to 750 C by casting on an iron plate, then reheated to 1000C at 5C per minute, soaked at 1000C for 3 hours and furnace cooled. The glass-ceramic product had a composition:
C~ponent Wt.
Na2O8.9 A1238.3 SiO248.7 TiO220.0 CaO14.1 and X-ray diffraction analysis showed sphene as the sole crystalline product.
Example 5 _,.
powdered mixture comprising 67.4 g Zeolon~ 900 (K-form) ~9.1 g Potassium titanate 23.2 g Calcium carbonate was heated in a ceramic crucible at 1400 C for 1 hour, cooled rapidly to 750C by removing the crucible from the urnace, then rehea`ed to 1000 C at 5C per minute, soaked at 1000C for 3 hours and then furnace cooled. The glass ceramic prodllct ha~l a composition:
C~mpor.ent Wt. %
K2O 14.2 A12O~ 6.9 Si2 40 7 rrio2 26.4 CaO 11.8 and X-ray ciiffraction analysis showed sphene as the sole crystalline product.
Example 6 A powdered mi~ture containing lant.anum was heated in a ceramic crucible at 1400C for 1 hour, cooled rapidly to 750C by removing the crucible from the furnace, then reheated to 1000C at 5C per minute, soaked at 1000C for 1 hour and then furnace cooled. The glass-ceramic prvduct had a composition:
Component Wt .
Na2O 5.84 CaO 11.61 TiO2 16.89 SiO2 48.92 l-la2o39.31 x-ray dif~raction analysis showed sphene as the sole 25 crystalline product. Scanning Auger microscopy has been used to distlnguish all of the lanthanum and has verified that it is contained substantially in the crystal phase.
Lanthanum in the glass was below the detection limit.
It will be appreciated that particular rererence has been made herein to the Na2O ~ Al.2O3 ~ SiO2 - Tio2 -CaO system as there is a high probability of Na2O in the waste materials anc'. an a]uminosili.cate glass has particularly advantageous properties. However, the present invention is not limited -to this system as it is possible to replace Na wi-th K cations, at least some of the Al with Fe and Cr (natural sphenes are known to contain Fe2O3, for example), at least some of the Ti with Zr, Sn, Zn, and the P with T.~ and Nb.
While particular reference has been made herein to inorganic ion exchange media, it will be appreciated that the present invention is not limited thereto, as organic ion exchange resins such as Duolite~ ARC 359 (Diamond Sham-rock) and Amberlite~ IRN-15Q or IRN 154 ~Rohm & Haas) may also be used in the production of glass-ceramic products.
Amberlite~ IRN-150, IRN-154 and IRN-300 are typically used to purify the moderator and primary heat transport systems of CANDU~ nuclear reactors. In exchange resins IRN-150, 154, 300 are mixed bed resins comprising both cation exchange resins and anion exchange resins. However, cation exchange resin alone could be operative.
Example 7 A mixture comprising:
30 g. Amberlite~ IRN-154 (Rohm & Haas) 8.4 g. Na2CO3 8.0 g. A12O3 28-Q cJ. SiO2 17.1 g. ~rio2 12.0 g. CaO

was heated in a ceramic crucible at 700C for 1 hour to incinerate the resin. The mixture of ash and chemicals was then heated at 1400C for 1 hour, cooled rapidly to 750 C by casting on an iron plate, then reheated to 1050 C
at 5C per minute, soaked at 1050C for 1 hour and furnace cooled. The glass-ceramic product had a composition:
Co~nent~t.~
_ _ _ Na2O 7.0 A123 11.5 Tio2 24.4 CaO 17.1 X-ray diffraction analysis showed sphene as the sole crystalline product.

Claims

WE CLAIM:

1. A process for preparing a glass-ceramic product for storing radioactive nuclear waste materials for extraordinarily long periods of time comprising:
(a) passing liquid radioactive waste materials through a selected ion exchange medium to thereby deposit said radioactive materials thereon and separate a relatively low level radioactive liquid for subsequent disposal;
(b) heating said ion exchange medium bearing said radioactive materials, with sufficient glass forming constituents being present, so as to form a melt, the non-radioactive portion of which has a composition in the range:
M2O 0 - 15 wt. %
M112O3 0 - 15 wt. %
SiO2 35 - 65 wt. %
TiO2 10 - 35 wt. %
M1O 0 - 15 wt. %
CaO 5 - 10 wt. %
M111O2 0 - 3 wt. %
M1V2O5 0 - 3 wt. %;

where M is selected from Na, K; M1 is selected from Ca, Ba, Sr; M11 is selected from Al, B, Fe, Cr; M111 is selected from Zr, Sn, Zn; and M1V is selected from P, Ta, Nb;
(c) cooling said melt so as to form a glass; and (d) heat treating said glass so as to crystallize sphene crystallites in a protective glassy matrix and containing said radioactive materials.

2. A process as claimed in claim 1 wherein said ion exchange medium is selected from a zeolite, titanate and calcium hydroxyapatite.

3. A process as claimed in claim 2 wherein said ion exchange medium bearing said radioactive materials is heated to a temperature in the range 1250 - 1600°C so as to form a melt.

4. A process as claimed in claim 3 wherein said glass is heated to a selected temperature above about 950°C but below the melting point of said glass.

5. A process as claimed in claim 4 wherein said glass is heated to a selected temperature in the range 950° -1050°C at a rate of 1-10°C per minute, soaked at said selected temperature for a period of at least 30 minutes so as to crystallize sphene crystallites in a matrix of glass, and furnace cooled.

6. A process as claimed in claim 5 wherein said sphene has a composition within the area a b c d in Figure 1.

7. A process as claimed in claim 2 wherein said zeolite is mordenite.

8. A process as claimed in claim 1 wherein said ion exchange medium comprises a cationic exchange resin.

9. A process as claimed in claim 8 wherein said exchange resin is a mixed bed cationic and anionic exchange resin.

10. A glass-ceramic product for storing radioactive nuclear waste materials for extraordinarily long periods of time, having a composition:
M2O 0 - 15 wt.%
M112O3 0 - 15 wt.%
SiO2 35 - 65 wt.%
TiO2 10 - 35 wt.%
M1O 0 - 15 wt.%
CaO 5 - 10 wt.%
M111O2 0 - 3 wt.%
M1v2O5 0 - 3 wt.%
where M is selected from Na, K; M1 is selected from Ca, Ba, Sr;
M11 is selec-ted from Al, B, Fe, Cr; M111 is selected from Zr, Sn, zn; and M1V is selected from P, Ta, Nb; and comprising sphene crystallites in a protective glassy matrix.

11. A glass-ceramic product as claimed in claim 10 having a composition in the range Na2O 6 - 10 wt.%
CaO 10 - 18 wt.%
TiO2 10 - 23 wt.%
A12O3 3 - 6 wt.%
SiO2 49 - 64 wt.%
P2O5 0.0 - 1.0 wt.%.

12. A glass-ceramic product as claimed in claim 11 having a composition 6.6% Na2O, 5.1% A12O3, 16.5% CaO, 14.8% TiO2 and 57% SiO2.

13. A cartridge for the treatment of liquid radioactive nuclear waste materials containing an ion exchange medium which, upon heating with sufficient glass forming ingredients forms a glassy product having a composition in the range M2O 0 - 15 wt.%
M11203 0 - 15 wt.%
SiO2 35 - 65 wt.%
TiO2 10 - 35 wt.~
M1o 0 - 15 wt.%
CaO 5 - 10 wt.3 M111O2 0 - 3 wt.%
M1V2O5 0 - 3 wt.%
where M is selected from Na, K; M1 is selected from Ca, Ba, Sr;
M11 is selected from Al, B, Fe, Cr; M111 is seleeted from Zr, Sn, Zn; and M1V is seleeted from P, Ta, Nb.

14. A cartridge as claimed in claim 13 wherein said ion exchange medium is selected from a zeolite, titanate, calcium hydroxyapatite, a cationic exchange resin and a mixed bed cationic and anionic exchange resin.

15. A cartridge as claimed in claim 14 wherein said zeolite is mordenite.

16. A cartridge as claimed in claim 13, 14 or 15 containing said ion exchange medium and an insoluble source of calcium which upon heating form said product.

17. A cartridge as claimed in claim 13, 14 or 15 wherein said product has a composition in the range Na2O 6 - 10 wt.%
CaO 10 - 18 wt.%
TiO2 10 - 23 wt.%
A12O3 3 - 6 wt.%
SiO2 49 - 64 wt.%
P2O5 0.0 - 1.0 wt.%.

18. A glass ceramic product as claimed in claim 10, 11 or 12 wherein at least a substantial proportion of said radioactive waste material is contained in said sphene crystallites.