Development of Ceramic Waste Forms for High-Level Nuclear Waste over the Last 30

Years

Eric Vance

Institute of Materials and Engineering Science, Australian Nuclear Science and Technology

Organisation, New Illawarra Road, Menai, NSW, 2234, Australia

ABSTRACT

Many types of ceramics have been put forward for immobilisation of high-level waste

(HLW) from reprocessing of nuclear power plant fuel or weapons production. After describing

some historical aspects of waste form research, the essential features of the chemical design and

processing of these different ceramic types will be discussed briefly. Given acceptable laboratory

and long-term predicted performance based on appropriately rigorous chemical design, the

important processing parameters are mostly waste loading, waste throughput, footprint, offgas

control/minimisation, and the need for secondary waste treatment. It is concluded that the

ìproblem of high-level nuclear wasteî is largely solved from a technical point of view, within the

current regulatory framework, and that the main remaining question is which technical

disposition method is optimum for a given waste.

INTRODUCTION

Solids for the immobilisation of high-level waste (HLW) from nuclear fuel reprocessing or

weapons production, have been under development for over 50 years. The principal aim is to

prepare a nearly water-insoluble solid, with minimal process wastes being produced. This can be

done by either (a) separating the radioactive ions from the bulk of the waste and then

incorporating them in the waste form or (b) incorporating the waste as a whole into a solid.

While method (b) involves fewer process steps in general, method (a) can lead to minimisation

of the waste form volume for HLW, and hence ease pressure on limited and very expensive

HLW repository space.

While it has been argued that the variability and predicted long-term uncertainties of waste

form performances are trivial with respect to the uncertainties in the near- and far-field predicted

performance of a geological repository, the fact that waste forms can be subjected to laboratory

examination remains a compelling driving force to persist with waste form performance studies.

Clearly improving waste form performance will decrease environmental risk irrespective of the

precise source term used in modelling.

As a rough guide, a waste form is considered as a serious candidate material if the

normalised leach rate of the most intrinsically soluble species (typically alkalis) in a

comparatively large volume (V) of hot deionised water is < 1 g/m

/day, based on geometrical

surface area (SA) and with SA/V being < 0.1 cm

-1

. The normalised leach rate pertains to the

fraction of the inventory leached out per unit time of a given species and not the absolute

quantity.

Immobilisation of HLW was seen as a problem as early as 1953 and a paper [1] dealt with

the use of fired clay for this purpose. Many types of waste forms-cements, glasses, glass-

ceramics, ceramics and metal alloys- have been subsequently proposed and researched. An

excellent history and technical description of waste form development up to the late 1980s

around the world is given in [2]. This includes detailed studies of spent fuel. However the

remainder of this paper will deal with ceramics and glass waste forms.

GLASS AND CERAMIC WASTE FORMS

From around 1960 to the mid-1970s, the then US Atomic Energy Commission developed

borosilicate glass to incorporate the waste fission products and actinides from reprocessing.

These were scaled up in the late 1960s to the full size dictated by the standard US disposal

canisters which are 3 m high x 0.61 m outer diameter. The concept was that the waste fission

products were generally soluble in such glass and the glass was reasonably leach-resistant. These

glasses can now be produced large melters having total throughputs of several T/day. The glass

is poured into the disposal canisters. The glasses can accommodate ~15 wt% of fission products

and actinides and are fairly resistant to leaching by typical groundwaters, such as would occur in

a deep geological repository.

In the mid-1970s Pennsylvania State University workers noted that borosilicate glasses were

fundamentally unstable from a thermodynamic point of view, especially in the presence of water,

and devised ceramic waste forms for HLW, based on the known natural longevity of crystalline

silicate, phosphate and molybdate minerals [3]. These "supercalcine" ceramics were sintered in

air at ~1100

C and had very high loadings of fission products, typically 70 wt% of fission

product oxides, and the chemistry of the different phases was driven by the fission products as

majority components. Typical phases were pollucite, CsAlSi

; powellite, CaMoO

; and rare

earth apatites and phosphates (e.g. monazite, (RE,An)PO

, where RE and An = rare earths and

actinides respectively). All of these had mineral analogues which were known to be very durable

in the hot, wet conditions likely to characterise a deep geological repository for the waste, and

had survived in nature for many millions of years.

Following work at Sandia Laboratories in the US on phase assemblages occurring on heating

sol-gel titania particles on which HLW fission products and actinides were sorbed [4], Ringwood

and his co-workers in Australia in the late 1970s devised multi-phase titanate ceramics in which

nearly all the fission products and actinides in HLW from nuclear fuel reprocessing were

incorporated substitutionally in the various phases [5].

Table I. Phase assemblage of synroc-C

Phase wt% Radionuclides in lattice

Hollandite, Ba(Al,Ti)

30 Cs, Rb

Zirconolite, CaZrTi

30 RE, An*

Perovskite, CaTiO

20 Sr, RE, An

Ti oxides 15

Alloy phases 5 Tc, Pd, Rh, Ru etc.

* RE, An = rare earths and actinides respectively,

Typical waste loadings were 20 wt% of HLW oxides and the production technology was the

addition of TiO

, ZrO

, CaO, BaO and Al

to the Purex-type HLW, calcination of the

waste/precursor mixture in a reducing atmosphere, followed by hot uniaxial pressing at ~1100

to make a dense ceramic. Table I gives the phase assemblage of synroc-C, the version which was

targeted to immobilise Purex-type reprocessing waste.

The principal advantage of this synroc (short for synthetic rock) ceramic was that the waste

ions were dilutely incorporated in durable titanate mineral phases which were considerably more

insoluble in water than the silicates and phosphates etc. used in supercalcine. Moreover, the

synroc waste loading could be varied between zero and 35 wt% using the same inert additive

chemistry without substantially changing the basic phase assemblage. This flexibility is a large

advantage.

Volatility losses of the supercalcine ceramics were severe, so the Rockwell Science Center in

California put forward hot isostatic pressing as the preferred consolidation method. The

Rockwell version of tailored ceramic for the (Al,Fe-rich) HLW at Savannah River National

Laboratory (SRL) in the early 1980s was based on a mixture of magnetoplumbite (CaAl

alumina, UO

and spinel [6] . Another type of synroc (synroc D) [7] was also devised for the

SRL waste. However after the decision in the US in 1981 to immobilise these wastes by

vitrification, very little effort on waste form development was made in the US until the mid-

1990s, although lead phosphate glass was studied for HLW immobilisation by Oak Ridge

National Laboratory researchers in the mid-1980s [8] and Catholic University workers continued

research on different varieties of silicate-based glasses [9]. Although the waste form properties of

the phosphate glasses were generally acceptable and the melting temperatures were low

(~800

C), refractory corrosion was an important problem. Also the use of lead per se was seen

as unattractive from the environmental point of view. The development of glass waste forms for

HLW continued in Europe, notably in France, and Canada began development of borosilicate

and aluminosilicate glasses and sphene glass-ceramics in the early 1980s.

Since the mid-1990s, workers at the University of Missouri have been studying iron

phosphate based glasses [10,11] which have low leach rates and little tendency for refractory

corrosion in Joule melters. Russian workers have used sodium aluminophosphate glass produced

by the cold-crucible melter (CCM) technique for some of their HLW [12]. The CCM [13] uses

higher temperatures than Joule melters and the use of a water-cooled crucible means that a

frozen glass layer lies between the melt and the metal crucible, inhibiting reaction between the

glass and the crucible. However volatile losses are potentially larger because of the higher

melting temperature. Radiofrequencies (rf) are used to carry out the heating and the coupling of

the rf to the charge is generally negligible in the cold state, so lumps of metal or graphite are

usually used as starters.

Immobilisation of actinides

The synroc family of ceramics have very fine grain sizes and excellent leaching behaviour,

up to 1000 times better than glass at extended leaching times. In the early 1990s, the synroc

ceramics were refocussed towards the study of zirconolite-rich materials [14] for immobilisation

of partitioned minor transuranic elements in advanced fuel cycles being studied in France and

Japan. The initial work during 1991-4 was directed at the latter application in conjunction with

the Japanese Atomic Energy Research Institute. Perovskite was also studied for comparison [15].

Work on surplus impure Pu immobilisation, with Lawrence Livermore National Laboratory

(LLNL) as the lead laboratory for the US Department of Energy (DOE), moved from zirconolite-

to pyrochlore-rich ceramics during 1994-7. This was because of solid solution limits in the first

instance when the target of the work changed from immobilisation of 10 wt% Pu alone to the

additional inclusion of 20 wt% U. Moreover, it was realised that anisotropic swelling effects

could potentially lead to microcracking problems after self-damage due to the alpha decay flux if

a non-cubic matrix was employed. The estimated time for amorphisation to be complete is on

the order of 1000 years and the resultant volume expansion would be around 6% [16]. These

ceramics incorporated an atom each of neutron-absorbing Gd and Hf for each atom of Pu to deal

with potential criticality in the sample [17]. Near-field aggregation of Pu due to leaching was

shown to be not a problem from the criticality aspect either, because the leach rates of Pu were

spanned by those of the neutron absorbers: hence any leached Pu would be accompanied by

neutron absorbers. The final baseline (no impurities) version of the ceramics chosen by the US

DOE in 1998 contained 95 wt% of a pyrochlore -structured Ca

0.89

0.23

0.44

0.22

phase plus 5 wt% of rutile-structured Ti

0.9

0.1

[17]. The form of the ceramic was to be 76

mm diameter pellets weighing ~500g and it was prepared by sintering at ~1350

C; sintering was

satisfactory as volatile losses were minimal and densities > 93% of theoretical could be obtained,

so that open porosity was negligible. This product was the first crystalline waste form to be

validated in the US. However in early 2002, it was decided to remove the disposal option for

US/Russian surplus Pu, and to proceed only with a mixed UO

-PuO

fuel option for

utilisation [16].

Notwithstanding this decision, many investigators have continued study of these materials

[18,19]. It has been pointed out that the abovementioned ceramics are prone to amorphisation by

alpha decay; although early work [20] suggested that aqueous dissolution rates would be greatly

increased as a result, more recent work [21] shows that amorphisation leads little if any increases

in dissolution rates. Also, while it has been shown that the substitution of Zr for Ti can lead to

greatly improved amorphisation resistance [18,22] , the range of impurities that can be

accommodated by the resultant fluorite structure is severely reduced, and the processing

temperature to assure densification is raised to ~ 1500

C [23].

Other ceramic research on waste forms to house minor actinides and rare earth fission

products is being conducted in Russia and France. Stefanovsky et al. are researching murataite

ceramics [24] and French workers have studied apatites [25] and more recently zirconolite glass-

ceramics [26].

DESIGN OF CERAMIC WASTE FORMS

Before discussing ceramics as such it is relevant to point out that while waste forms which

are fabricated by relatively simple and cheap low-temperature cementation or spray processing,

these materials have open porosity in the case of cement and may be produced only as fine

powders in the latter case, Even though these low-temperature products can formally be

regarded as ceramics and these materials can pass tests such as the PCT-B protocols [27],

especially if BET surface areas are used in determining leach rates, there is no doubt that for

valid comparison with the more traditional glasses and dense ceramics, the lack of densification

in the low-temperature products militates against their use and such products will not be further

considered here. If powders are used, irrespective of whether BET areas are used, the mass loss

by dissolution of the given elements in a test should be specified. Moreover the PCT test can

give rise to saturation effects, and if used alone can give a very misleading picture of leach

resistance-MCC-1 type tests would give a more realistic picture [28].

Design of ceramic waste forms for HLW requires an integrated approach to chemical design

and processing method. It is important to realise that in ceramic design the waste ions are

substituted for host ions, unlike for glass in which the waste ions are merely added to a

precursor. When the waste ions in the ceramic phases have a different chemical valence than

the host ions, lattice defects which provide charge compensation can form. To understand all the

design factors means that we need our strong underlying science program, directed at crystal

chemistry, aqueous dissolution and radiation damage effects. Processing will need

understanding of mixing the waste with selected additives, calcination, and hot consolidation,

noting that the processing technique influences the final phase assemblage, i.e. chemical design

and processing have to be conducted in an integrated fashion.

Single-phase ceramics have been widely studied from a scientific point of view for both

single radioactive elements formed by partitioning of reprocessing wastes or the entire

complement of elements in a given HLW. However in actual production, it would be mandatory

to not have to rely on an exact match of waste and precursor masses in multi-cation hosts, such

as those mentioned above, as such an exact match is industrially unrealistic, especially as wastes

are by no means homogeneous. So what is needed is an "extra" phase(s) whose abundance may

vary as the waste/precursor ratio or the waste chemistry varies, while still maintaining the same

qualitative phase assemblage-as in the synroc-type ceramics (see above) in which rutile is the

extra phase. Therefore in multi-phase ceramics designed to immobilise different waste ions in

different phases, a key factor is the compatibility of the target phases.

A variety of glass-ceramics in which slow-cooling or reheating to intermediate temperatures

of melts produces crystalline phases in a glassy matrix have also been widely studied. The idea

is to partition the waste radionuclides into the more durable crystalline phases while retaining

melting technology for high throughputs. Early work centred around the AECL sphene glass-

ceramic [29] and Idaho basalt-glass ceramics [30] in the 1980s, and more recent work has been

by French (see above) and ANSTO workers (see below).

Between 1994 and 1997 a parallel effort was developed at ANSTO to study glass-ceramics

for immobilisation of Hanford tank wastes (WA, USA). The object of this work was to contain

as much as possible of the actinides in synroc phases, principally zirconolite, in a boroaluminate

silicate glass matrix. Other crystalline phases such as zirconia and CaF

were also present.

These glass-ceramics had waste loadings of 50-70 wt% and leach rates were often 10-100 lower

than those for the regulatory reference material for the aqueous durability of waste forms. This

ANSTO effort has been refocussed since 2000 and a variety of glass-ceramics have been devised

for liquid HLW and HLW calcines located at the Idaho National Laboratory [31], as well as Pu-

bearing wastes in Britain [32].

Processing technology

The disadvantages of Joule melters is that they have finite lifetimes due to refractory

corrosion and the temperatures cannot exceed ~1150

C. Also, the footprint associated with the

offgas systems are inevitably large, and failed melters constitute large amounts of secondary

radioactive waste.

For radioactive waste, the calcined waste form material is placed inside a sealed metal can

which is consolidated to full density by heating and compressing it with high-pressure ( ~100

MPa) argon gas. The use of metal inhibits the reaction between the ceramic waste form and the

container [33] and of course prevents offgas escape. So the entire process produces offgas only

in the calcination stage where temperatures are much lower than those in the hot-pressing. Hot

isostatic pressing in waste form production has been validated recently at Argonne-West by US

regulators. In addition, the choice of a suitable can material of sufficient thickness can provide

immobilisation by this technology.

Influence of AFCI /GNEP

The main influence of AFCI/GNEP is that minor actinides generated in nuclear fuel and

which were previously considered as waste would be considered as resources to be burnt in fast

reactors. Though complete burning of such actinides is difficult as the conversion efficiency is

proportional to the actinide concentration in the fuel, an overall reduction in the waste actinide

inventory relative to the fission product inventory will reduce the mean half-life of the waste,

ideally from millions of years to hundreds. Thus there is increasing interest in recycling used

nuclear fuel, especially in pyroprocessing using electrolysis of fuel dissolved in molten alkali

halide salts to separate out actinides from fission products.

FUTURE OUTLOOK

After 50 years of research worldwide on all relevant aspects of different waste forms-waste

loading, leach resistance in deionised and likely repository-type groundwaters, self-irradiation,

full-scale active demonstrations etc, only a small percentage of high-level radioactive waste

(other than spent fuel) around the world has been converted to a qualified waste form. Moreover

virtually none has been actually put into a geological repository. In summary this situation has

arisen because political acceptance of such repositories is limited (ìnot in my backyardî

syndrome), but also because waste form choices are still a matter of considerable debate,

especially by the proponents of the different alternatives. In addition, because different legacy

HLWs are chemically different, the optimum waste forms may very well be different. Further,

regulatory requirements are different in different countries notwithstanding the umbrella position

put forward by the International Atomic Energy Agency. And because HLW is not a product per

se, it is always cheaper to postpone disposal as long as there are no current problems with

leakage of tanks etc.

But it is a positive feature that the differences between the performances of vitreous and

alternative (including ceramics) waste forms are narrowing, with both improving. Moeover, it

can now be argued that ìthe problem of HLWî is essentially solved from a technical point of

view within the current regulatory framework and that the ultimate engineering means of actual

safe, economical high-level waste disposition will come down to processing advantages for

particular types of HLW.

ACKNOWLEDGEMENT

I wish to acknowledge here the collaborative efforts of numerous colleagues at ANSTO and

around the world over many years.

REFERENCES

1. L. P. Hatch, American Scientist, 41, 410 (1953).

2. Radioactive Waste Forms for the Future, edited by W. Lutze and R. C. Ewing (North-

Holland, Amsterdam, 1988).

3. G. J. McCarthy, Nucl. Tech., 32, 92 (1977).

4. R. G. Dosch, A. W. Lynch, T. J. Headley and P. Hlava, in Scientific Basis for Nuclear Waste

Management, Volume 3, edited by J. G. Moore (Plenum Press, New York, 1980) pp. 123-30.

5. A. E. Ringwood, S. E. Kesson, N. G. Ware, W. Hibberson, A. Major, Nature (London), 278,

219 (1979).

6. P. D. E. Morgan, D. R. Clarke, C. M. Jantzen and A. B. Harker, J. Amer. Ceram. Soc., 64,

249 (1981).

7. A.E. Ringwood, S. E. Kesson and N. G. Ware, in Scientific Basis for Nuclear Waste

Management, Volume 2, edited by C. J. M. Northrup (Plenum Press, New York, 1980) pp.

265-73.

8. B. C. Sales and L. A. Boatner, in Ref. 2, pp. 193-231.

9. I. S. Muller, A. C. Buechele, I. L. Pegg and P. B. Macedo, in Scientific Basis for Nuclear

Waste Management XV, edited by C. Sombret (Materials Research Society, Pittsburgh, PA,

USA, 1992) pp. 91-8.

10. G. K. Marasinghe, M. Karabulat, X. Fang, C. S. Ray, D. E. Day, D. L. Caulder, J. J. Bucher,

N. M. Edelstein, D. K. Shuh, and P. G. Allen, in Environmental Issues and Waste

Management Technologies in the Ceramic and Nuclear Industries V, edited by G. T.

Chandler and X. Feng (American Ceramic Society, Westerville, OH, USA, 2000) pp. 115-

122.

11. C-W. Kim, D. Zhu and D. E. Day, in Environmental Issues and Waste Management

Technologies in the Ceramic and Nuclear Industries VIII, edited by S. K. Sundaram, D. R.

Spearing and J. D. Vienna (American Ceramic Society, Westerville, OH, USA, 2003) pp.

329-36.

12. A. P. Mukhamet-Galeyev, L. O. Magazina, K. A. Levin, N. D. Samotoin, A. V. Zotov and B.

I. Omelianenko, in Scientific Basis for Nuclear Waste Management XVIII, edited by T.

Murakami and R. C. Ewing (Materials Research Society, Pittsburgh, PA, USA, 1995) pp. 79-

86.

13. A.V. Demine, N. Y. Krylova, P. P. Poluektov, I. N. Shestoporov, T. Y. Smelova, V. F. Gorn

and G. M. Medvedev, in Scientific Basis for Nuclear Waste Management XXIV, Eds. K. P.

Hart and G. R. Lumpkin (Materials Research Society, Warrendale, PA, USA, 2001) pp. 27-

33.

14. E. R. Vance, B. D. Begg, R. A. Day and C. J. Ball, in Scientific Basis for Nuclear Waste

Management XVIII, Eds. T. Murakami and R. C. Ewing (Materials Research Society,

Pittsburgh, PA, USA, 1995) pp. 767-774.

15. H. Mitamura, S. Matsumoto, T. Tsuboi, E. R. Vance and B. D. Begg, ibid., pp.1405-12.

16. R. C. Ewing, W. J. Weber and J. Lian, J. Appl. Phys., 95, 5949-71 (2004).

17. E. R. Vance, A. Jostsons, S. Moricca, M. W. A. Stewart, R. A. Day, B. D. Begg, M. J.

Hambley, K. P. Hart and B. B. Ebbinghaus, in Environmental Issues and Waste Management

Technologies in the Ceramic and Nuclear Industries IV, edited by J. C. Marra and G. T.

Chandler, (American Ceramic Society, OH, USA, 1999) pp. 323-9.

18. K. E. Sickafus, L. Minervini, R. W. Grimes, J. A. Valdez, M. Ishimura, F. Li, K. J.

McClellan and T. Hartmann, Science

, 289, 748 (2000).

19. J. Lian, J. Chen, L. M. Wang, R. C. Ewing, J. M. Farmer, L. A. Boatner and K. B. Helean,

Phys. Rev. B 68, 134107 (2003).

20. J. W. Wald and W. J. Weber, in Advances in Ceramics, Vol 8, edited by G. G. Wicks and W.

A. Ross (American Ceramic Society, Columbus, OH, USA, 1984) pp. 71-5.

21. D.M. Strachan, R.D. Scheele, A.E. Kozelisky and R.L. Sell, Pacific Northwest National

Laboratory Report, PNNL-14232 (2003).

22. B. D. Begg, N. J. Hess, D. E. McCready, S. Thevuthsan and W. J. Weber, J. Nucl. Mater.,

289, 189 (2001).

23. M. W. A. Stewart, B. D. Begg, E. R. Vance, M. Colella, K. Finnie, K. P. Hart, H. Li, G. R.

Lumpkin, K. L. Smith and W. J. Weber, in Scientific Basis for Nuclear Waste Management

XXV, edited by B. P. McGrail and G. A. Cragnolino (Materials Research Society,

Warrendale, PA, USA, 2002) pp. 311-8 .

24. S. V. Yudintsev, S. V. Stefanovsky, B. S. Nikonov and B. I. Omelianenko, in Scientific Basis

for Nuclear Waste Management XXIV, edited by K. P. Hart and G. R. Lumpkin (Materials

Research Society, Warrendale, PA, USA, 2001) pp. 357-65.

25. L. Boyer, J. Carpena, J. L. Lacout, Solid State Ionics, 95, 121 (1997).

26. D. Caurant, O. Majerus, P. Loiseau, I. Bardez, N. Baffier and J. L. Dussossoy, J. Nucl.

Mater., 354, 143 (2006).

27. Determining Chemical Durability of Nuclear Hazardous and Mixed Waste Glasses: The

Product Consistency Test (PCT), ASTM Designation: C1285-97.

28. M. L. Carter, this meeting.

29. P. J. Hayward, in Ref. 2, pp. 427-93.

30. P. V. Kelsey, R. P. Shuman, J. M. Welch, D. E. Owen and J. E. Flinn, in Scientific Basis for

Nuclear Waste Management, Vol 6, edited by S. V. Topp (North-Holland, Amsterdam, 1982)

pp. 533-40.

31. R. A. Day, B. D. Begg, S. Moricca, M. W. Stewart, R. Muir and E.R. Vance, WMí05,

Tucson, AZ, USA, CD-ROM (2005).

32. C.R. Scales, E.R. Maddrell, N.Gawthorpe, B.D. Begg, S. Moricca, R.A. Day and M.W.A.

Stewart, WMí06, Tucson, AZ, USA, CD-ROM (2006).

33. H. Li, Y. Zhang. P. J. McGlinn, S. Moricca. B. D. Begg and E. R. Vance, J. Nucl. Mater.,

355, 136 (2006).