Ceramic Waste Forms

5.19

Ceramic Waste Forms

E. R. Vance Institute of Materials Engineering, Australian Nuclear and Technology Organisation, Menai, NSW, Australia

Crown Copyright ß 2012 Published by Elsevier Ltd. All rights reserved.

5.19.1

Introduction

485

5.19.2 5.19.3 5.19.4 5.19.5 5.19.6 5.19.7 5.19.8 5.19.9 5.19.10 5.19.10.1 5.19.10.2 5.19.10.3 5.19.11 5.19.12 References

Desirable Performance Characteristics of High-Level Nuclear Waste Forms Design of Waste Form Ceramics Historical Evolution of Candidate Ceramic Waste Forms for HLW Titanate Ceramics Glass–Ceramics Aqueous Dissolution Radiation Damage Thermodynamic Stability of Multication Oxides Processing of Ceramics and Glass–Ceramics Hot Uniaxial Pressing Hot Isostatic Pressing Melting Cements and Geopolymers Conclusions

488 488 490 492 494 495 495 496 496 497 497 498 499 500 501

Abbreviations An ANSTO

Actinide Australian Nuclear Science and Technology Organisation AWE Atomic Weapons Establishment BET Brunauer–Emmett–Teller DWPF Defense Waste Processing Facility EA Environmental Assessment FP Fission product FUETAP Formed under extreme temperatures and pressures HIP Hot isostatic press HLW High-level waste HUP Hot Uniaxial Press ICPP Idaho Chemical Processing Plant ILW Intermediate-level waste LLNL Lawrence Livermore National Laboratory LLW Low-level waste MCC Materials Characterization Center MOX Mixed oxide MPP Magnesium potassium phosphate NIMBY Not in my Backyard NZP Sodium zirconium phosphate OPC ordinary Portland cement PCT Product consistency test PNNL Pacific Northwest National Laboratory RE Rare earth

SI SRL Synroc USDOE

International System of Units Savannah River National Laboratory Synthetic rock United States Department of Energy

5.19.1 Introduction Worldwide inventories of high-level waste (HLW) from nuclear power and weapons production, other than spent fuel, constitute hundreds of thousands and perhaps millions of tonnes.1 Figures 1 and 2 show the abundances and the time dependences of the radioactivity of HLW from nuclear fuel reprocessing, respectively. Disposal of HLW other than spent fuel itself basically involves adding specific material to the waste and processing it to transform it to a dense refractory solid – the waste form – that can withstand prolonged immersion in groundwater. In a large majority of cases, the solid would be an oxide, but metallic components may also be present, depending on the waste and the processing conditions. The solid, which would be a glass, ceramic, glass–ceramic, or cementitious material, depending on the nature of the waste to be dealt with, would then be transferred in a metal disposal canister to a geological repository (Figure 3), in which the radioactive and toxic 485

486

Ceramic Waste Forms

10

14 Mev 103

1 Fission product total activity

Fission yield (%)

102 0.1

10 Actinides total activity

1 0.01 10−1

Thermal 0.001

0.0001 70

10−2

10−3 90

110 130 Mass number

150 10−4

Figure 1 Relative atomic abundances of fission products elements in a commercial power plant nuclear fuel.

10−5

components would be immobilized for long periods, up to millions of years if necessary, depending on the half-lives of the dangerous species. Glass is currently seen as a baseline option for HLW. In this article, the focus of the immobilization is on the waste form, but of course the repository would also be expected to play a significant role in preventing the transport of radionuclides and toxic ions to the biosphere. Because it seems unlikely that spent fuel in the form of irradiated UO2, a ceramic, will be chemically processed, but rather simply allowed to cool as the short-lived fission products decay, and then containerized for eventual repository disposal, we will not discuss spent fuel in this article, but it will be the subject of Chapter 5.16, Spent Fuel as Waste Material. Tables 1 and 2 show approximate compositions of Purex-type reprocessing and tank wastes in the United States. Since the activity of the waste falls with increasing time, it is technically advantageous to store the waste as long as possible, although this should not be seen as an excuse to delay immobilization unduly. Moreover, the method of storage is critical and needs frequent attention. For instance, a strong initial driver of HLW cleanup in the United States in the early 1990s was that at the Hanford reservation in the state of Washington, the stainless steel tanks containing the old military wastes from nuclear weapons production were leaking into the

10−6

1

10

102 103 104 105 106 107 Time (year)

Figure 2 Time dependence of radioactivity of reprocessing waste from commercial power plant nuclear fuel.

Earth’s surface

Rock Backfill, plus cement

1–3 km

Canisterized HLW Figure 3 Schematic diagram of a deep geological repository.

surrounding environment. Also, the tanks in which the water had largely evaporated over the years of storage, because of radiogenic heating, gave periodic gas evolution in the form of large hydrogen bubbles,

Ceramic Waste Forms

which had safety implications via potential radionuclide removal from the tank into the atmosphere and ground area adjacent to the tanks, as well as ignition and explosion. Historically, there has been explicit and implicit competition between ceramics and borosilicate glasses for immobilization of HLW (e.g., the ‘Atlanta shootout’) for Savannah River Laboratory (SRL) waste in 19812,3 and the US Department of Energy (USDOE) decision on impure US/Russian surplus Pu in 1998.4,5 However, as there are now so many types of HLW or intermediate-level wastes (ILWs) having a large variety of chemical compositions remaining to be immobilized around the world, it is becoming clear that glass, glass–ceramics, ceramics, and on occasion, cementitious materials have their place for different HLWs and ILWs. In this article, the focus is on ceramics and glass– ceramics targeted mainly to immobilize HLW.

Table 1 Approximate compositions (wt%) and halflivesa of main fission product (FP) and actinide oxides in Purex fuel reprocessing HLW that has been stored for >10 years FP oxideb

wt%

Half-life (year)

Cs2O SrO BaO RE2O3 ZrO2 MoO3 TcO2 AnO2 RuO2 PdO Rh2O3

6 3 4 15 15 15 6 6 10 6 2

30 30 – 100a – – 210 000 >10 000 – – –

Water excluded; RE: rare earth; An: actinide. a Group half-lives are very approximate as they range from short to long times for different components. Absence of half-life value ¼ stable elements. b Contains additional stainless steel corrosion products.

487

The specific activity of reprocessing HLW is in the order of tens of terabecquerels per liter, while the US tank wastes from Pu production for atomic weapons production have specific activities that are perhaps 1000 times lower. Indeed, in other countries, some of these latter wastes would be considered as ILW. We note that hot isostatic pressing (HIPing) in waste form production has been validated recently at INL, ID, USA (Idaho National Laboratory) by US regulators, and a de facto validation of crystalline waste forms may be taken from the 1998 decision,4 albeit overturned in 2001 in the sense of not proceeding with the immobilization option, to use ceramics for surplus Pu disposition in the United States. Ceramic materials for the incorporation and the immobilization of nuclear waste range from refractory, dense, fine-grained ceramic, or glass–ceramic candidates for HLW to cementitious product options for low-level waste (LLW). Here, we discuss the design of mainly ceramics and glass–ceramics for HLW, but we also discuss cementitious materials for ILW. How such materials meet the regulatory criteria for properties such as retention of radionuclides and toxic ions when the materials are exposed to water, fire resistance, and mechanical strength, together with other properties such as waste loading, and simplicity or otherwise of processing methods, will be outlined. The influence of self-damage due to decay of the radioactive species is also an important feature and is briefly mentioned, noting that detailed discussions of this phenomenon appear in Chapter 5.22, Minerals and Natural Analogues. The final destination of HLW is generally agreed to be a deep (0.5–3 km) geological repository (Figure 3), so the principal object of fabricating a waste form for HLW is, using the simplest and cheapest possible processing methods, to incorporate the waste in a solid that is minimally porous and has high leaching resistance when exposed to water, adequate strength, and a high fraction (>20 wt%) of waste per

Table 2 Approximate oxide compositions and half-lives of An, FP, and process chemical oxides in US tank wastes and surplus Pu-bearing waste (ignoring minor cations and assuming anions such as nitrates, nitrites, hydroxides, and carbonates are removed upon calcination) US tank waste (wt%)

Half-life (year)

Surplus Pu-bearing waste (wt%)

Half-life (year)

Na2O (40) K2O (5) Al2O3 (50) Fe2O3 (4) FP (<1) AnO2 (<1)

– – – – various >10 000

PuO2 (20) UO2 (20) Fe2O3 (20) Al2O3 (20) CaF2 (20)

24 000 4  109 – – –

An: actinide.

488

Ceramic Waste Forms

unit volume. A limiting feature for HLW is the radiogenic heat output (see, e.g., Sizgek6) as it is desirable to keep the repository temperature to 100  C or less to minimize the reactivity of the waste form with groundwater. After presenting some desirable characteristics of waste forms and features impacting on the design of multiphase ceramics for wastes containing many radionuclides and process chemicals, we outline the historical development of ceramic waste form research. Alternative means of preparing appropriate candidate waste form ceramics is put forward. Then, we conclude that the best way to treat a given nuclear waste depends on the nature of the waste itself and that there is no single optimum method of treating all nuclear wastes. We further note that disagreement among waste form proponents as to the best way to dispose of a given waste does not suggest that the ‘nuclear waste problem’ is not basically solved. Given this, it would follow that nuclear power is sustainable from that point of view. In the design of an appropriate ceramic, it is advantageous if the component phases are similar to those of naturally occurring multication oxide minerals such as oxyapatite (Ca2La8(SiO4)6O2) or zirconolite (CaZrTi2O7) that are known to survive in hot, wet conditions for millions of years. This feature and the relative thermodynamic stability of crystalline material versus amorphous glass were the principal stimuli in the 1970s for the candidacy of ceramic waste forms for HLW.

5.19.2 Desirable Performance Characteristics of High-Level Nuclear Waste Forms Again, it is imperative that waste forms have very high chemical durabilities in terms of resistance to leaching by groundwater. The durability of the waste form can be subjected to laboratory study and then optimized. In this context, normalized leach rates of <1 g m2 day1 at temperatures below 100  C are considered as baseline. One gram per square meter per day (1.16  105 g m2 s1) corresponds to around 0.2–0.5 mm day1 in terms of thickness. (While g m2 day1 is not an SI unit, it is almost universally used within the nuclear waste community.) The normalized leach rate is defined as the gross release rate of the ion in question divided by the concentration of the ion, and the aim of using normalization was to defy the ‘dilute and disperse’

strategy in which sufficient dilution could reduce gross leach rates to apparently acceptable values. The latter approach is objectionable because it creates very large volumes of (weakly) radioactive waste, which take up space and inflate cleanup costs. The effect of self-irradiation needs to be taken into account in these determinations. The higher the proportion of waste that can be incorporated per unit volume of the waste form, the less repository space will be needed and so the costs will be minimized. The waste form needs to be easily and reliably processed in a remote environment, and minimization of secondary wastes such as radioactive off-gases (needing recycle) during the production of the waste form is important. Given the need for aqueous durability, it is obvious that open porosity would be a very bad feature because entry of water into the interior of the waste form solid greatly increases the potential for leaching. This is a key factor in fabrication of waste forms and is discussed later in regard to cementitious materials.

5.19.3 Design of Waste Form Ceramics When waste ions, assumed to be present as oxides or as compounds that form oxides upon melting in air, are incorporated in borosilicate glass, they are normally incorporated as network formers or network modifiers, and the waste oxides can simply be added to the precursor glass chemicals, which are usually in the form of glass frit produced by pouring the molten glass into water. However, the situation is somewhat different when waste oxides are added to targeted ceramic precursors with the object of forming nearly water-insoluble multication mineral oxide phases, insofar as the waste ions in the mineral phases enter by a substitution mechanism, not by simple addition. The partitioning of the waste ions will depend on the cation site in which they substitute so that unequal partitioning of the waste ions will result in the formation of extra phase(s) if the waste ions are just added to the mineral phase precursor. Therefore, substitution of waste ions in a multication mineral phase requires prior adjustment of the overall stoichiometry and detailed knowledge of the preferred sites of the waste ions in the mineral phase. Moreover, if the valences of the guest and host ions are different, charge compensation is necessary to preserve electroneutrality of the phase. If the valence of the guest ion is less than that of the host ion, charge compensation can be

Ceramic Waste Forms

maintained by oxygen vacancies or by the addition of cations with higher valences than that of the host. So, for example, if it is desired to substitute two monovalent alkali ions into two divalent host ion sites, we can maintain phase electroneutrality by 2M2þ þ 2O2 $ 2Mþ þ O2 þ hO

½I

□O is a vacant oxygen site. However, we can also introduce a trivalent charge compensating cation for a monovalent ion substituted in divalent host sites via 2M2þ $ Mþ þ M3þ

½II

Similar considerations arise when it is desired to incorporate higher valence guest ions on to host sites, except that the charge compensators will be cation vacancies or substitutional cations with valences less than those of the host ions. Ceramics containing several phases are necessary when a full range of fission products, minor actinides, and process chemicals require immobilization, and the same principles of chemical accounting apply to multiphase ceramics made up of mutually compatible mineral phases. A first guess as to the way that the waste ions will incorporate themselves in the mineral phases will derive from a similarity of the approximately known7 ionic sizes between the host sites and the waste ions. The ionic sizes depend also on the valence of the ions so that the valences of the waste (and the host) ions need to be known. Also, the more similar the valences of the guest and host ions, generally the larger the solid solubility of the guest ions in the host phase. The valence depends on the crystal chemistry of the host site as well as on the prevailing oxygen fugacity (note that some elements such as the Pd group will likely form metals under some processing conditions, even oxidizing environments). A good example of crystal-chemical valence stabilization is the fact that monazite, CePO4, in which Ce exists as Ce3þ, can be fabricated by firing in air at 1200  C, whereas CeO2 (Ce4þ) can also be similarly formed in air, and analogous results are obtained for PuPO(48) and PuO2. Table 3 is a partial ‘library’ of candidate ceramic phases for radionuclide immobilization. Experimental studies to define the valences of variable-valence ions fortunately do not require the use of radioactive ions if stable ions of the corresponding elements exist and can be made from X-ray near-edge absorption spectroscopy, electronic optical spectroscopy, electron paramagnetic resonance,

Table 3

489

Partial ‘library’ of ceramic host phases

Phase

Radionuclide

Pollucite, CsAlSi2O6 Hollandite, (Cs,Sr,Ba,Rb)1.14 (Al,Ti3þ,Fe)2.28TiO16 Feldspar, CaAl2Si2O8 Apatite, Ca10([P,Si]O4)6(OH,F,Cl)2 NZP (Na,Ca0.5)(Zr,Ti)2(PO4)3 Monazite, REPO4 Garnet, Ca1.5GdTh0.5FeFe3SiO12 Zircon, ZrSiO4 Xenotime, YPO4 Zirconolite, CaZrTi2O7 Perovskite, CaTiO3 Fluorite, (RE,An)O2 Pyrochlore, RE2Ti2O7 Titanite, CaTiSiO5 Rutile, TiO2 Sodalite, Na4Al3Si3O12I

Cs Cs, Rb, Sr, Ba Sr, Ba RE, An, Sr, Ba Many RE, An RE, An RE, An RE, An RE, An Sr, RE, Tc, An RE, An RE, Zr, An RE, An, Sr Tc I

RE: rare earth; An: actinide; NZP: sodium zirconium phosphate.

electronic energy-loss spectroscopy, and Mossbauer experiments, to name but a few. Scanning/transmission electron microscopy can be employed to derive the stoichiometry of the various phases. For actinides and Tc, the actual radioactive isotopes have to be employed, although popular inactive simulated substitutes are Ru for Tc and Ce for the trivalent and tetravalent actinides. Facilities for examination of significantly radioactive samples are restricted to a few national laboratories in various countries. Having established the relevant ionic valences and the desired set of candidate mineral analog phases in the multiphase ceramics, the next requirement is to establish that the different phases are compatible (can coexist at elevated temperatures). Although attempts have been made to construct single phases to immobilize single or a range of waste ions, this strategy is difficult (see Section 5.19.4). Also, for a given waste form for a given HLW chemical composition, it is important that the waste form properties are flexible and not unduly compromised because of mismatches of waste/additives ratios, and variations of waste form chemistry, noting that HLWs are very frequently inhomogeneous mixtures of solutions and sludges, calcines, etc. Flexibility derives from the use of multiple phases and chemical buffering via the presence of a phase(s) that does not include radionuclides – then variations of chemical composition just result in a change of the proportions of the phases present, not the identities of the phases themselves.

490

Ceramic Waste Forms

5.19.4 Historical Evolution of Candidate Ceramic Waste Forms for HLW Although borosilicate glass worldwide had been the main candidate waste form for Purex-type HLW (see Chapter 5.18, Waste Glass) up to the mid-1970s, Pennsylvania State University workers noted9 that glasses were fundamentally unstable from a thermodynamic point of view, and they devised ceramic waste forms for HLW derived from nuclear fuel reprocessing, based on the known natural longevity of crystalline silicate, phosphate, and molybdate minerals. These so-called supercalcine ceramics10 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 (CsAlSi2O6), powellite (CaMoO4), and rare earth apatites (Ca2RE8(PO4)6O2) and phosphates (e.g., monazite, REPO4) (RE is trivalent rare earth). All of these had mineral analogs that were known to be very durable in the hot, wet conditions likely to characterize a deep geological repository for the waste. Following work at Sandia Laboratories in the United States on phase assemblages occurring on heating sol-gel titania particles on which simulated HLW fission products and actinides were sorbed,11 Ringwood and his coworkers in Australia in the late 1970s devised multiphase titanate-based ceramics in which nearly all the fission products and actinides in HLW from nuclear fuel reprocessing were incorporated substitutionally in the various mineral analog phases.12–14 Typical waste loadings were 20 wt% of HLW oxides and the production technology was slurry mixing of the waste and precursor oxides, calcination of the waste/precursor mixture in a reducing atmosphere, followed by hot uniaxial pressing (HUP) at 1100  C to make a dense ceramic. These ceramics are discussed in a little more detail in following sections. Primarily, these ceramics were focused on Purex-type HLW, but some work was done to immobilize SRL tank wastes from weapon production. At about the same time, and perhaps in part driven by the appearance of the synroc-type ceramics, there was a worldwide surge of interest in the immobilization of HLW. However, in the United States, a key decision was made during 1981–1982 to use borosilicate glass to immobilize HLW at SRL,2,3 and there was a substantial decrease in US funding for HLW waste form

research from then on. Nevertheless, a variety of alternative waste form development work continued around the world, and the book by Ewing and Lutze15 gives an excellent survey of research up to nearly the end of the 1980s. Candidate materials included glasses, ceramics, glass–ceramics, cermets, coated materials, and cements. However, in the course of time, it has been widely (but not universally) agreed that the only real remaining candidate types of material for HLW immobilization are glasses, ceramics, and glass–ceramics. These can be produced by Joule or cold-crucible melters, sintering or hot pressing, and particularly HIPing. However, cementitious materials may yet be useful for less active HLW such as US defense wastes from weapons production. Waste form development for HLW is still continuing in some shape or form in different nuclear countries, although Japan chose borosilicate glass in the mid-1990s and therefore ceased work on alternatives except in some niche areas, such as immobilization of 129I. France instituted the ‘law of 1991,’ which placed a moratorium on waste disposal until 2006, giving them 15 years of research to make a decision on the best choices of waste forms for their particular HLWs. In the 1990s, much work was done in France on apatites as candidate ceramics, and collaborations with the ANSTO group on titanate ceramics were initiated. A Th double phosphate Th4(PO4)4P2O7 ceramic for immobilization of tetravalent actinides has also been under development in France for several years. The material is claimed to have good aqueous chemical durability even when amorphized by swift heavy ions.16 Although numerous scientific papers on candidate ceramic waste forms still emanate from France, it is as yet not totally clear whether France will ultimately continue only with vitrification for their HLW or whether ceramics may play some role. Work has continued in Australia on titanate ceramics (see Section 5.19.5) although since around 2000, the work has evolved from a narrow focus on titanate ceramics to the use of HIPing technology to fabricate candidate glass–ceramic and ceramic waste forms to immobilize a wide range of HLW/ILWs that are generally perceived to be not very suitable for vitrification. Such waste forms have been directed at Pu-bearing wastes, Magnox sludges (UK), K-basin sludges (Hanford, WA, USA), Tc-rich waste,17 I-rich waste,18 Cs/Sr/Ba/Rb partitioned wastes,19 pyroprocessing wastes,20 and U-rich 99Mo production wastes.21 Collaborative work has also been done with UK workers on titanate ceramics, and UK workers have also

Ceramic Waste Forms

worked on other ceramic formulations. The AWE plc (Aldermaston, UK) has had an ongoing 10-year program to immobilize ILW derived from pyrochemical processing of Pu metal: they are using chlorapatiteand spodiosite (Ca2(PO4)Cl)-based ceramics.22 At ANSTO, work is being carried out under the Europart effort on pyroprocessing to try to immobilize waste pyroprocessing chloride salt in apatite-based ceramics.20 Russian workers are investigating murataite and garnet-based ceramics, together with perovskites.23,24 Also in Russia, investigations of the crystal chemistry of sodium zirconium phosphate (NZP)structured and other phosphates having a wide range of ionic substitutions have been carried out by Orlova’s group for many years.25 The effects on the NZP structure of a-damage from 238Pu and 239Pu substitutions have been observed.26 Other phosphates, notably monazites, have also been studied in regard to a-damage effects from 238Pu doping.27 It is appropriate at this stage to reiterate the diverse nature of HLW, depending in part on whether it derives from commercial Purex-type reprocessing or military Pu production. Generally speaking, the Purex wastes are highly active and are basically solutions of fission products and minor actinides in 3 mol l1 nitric acid solutions. The Pu production tank wastes, however, consist mainly of process chemicals, a relatively small amount of fission products in an alkaline solution derived from the need to neutralize nitric acid that would in time attack the stainless steel tanks. Thus, these wastes consist of a concentrated solution of Na salts plus hydroxide-rich sludges and are very inhomogeneous (and largely uncharacterized) even in single tanks. Hence, there is a possible need to actually separate individual wastes into solution and sludge fractions, and a definite need to design waste forms that can cope with diversity and compositional uncertainty. More importantly, longer-term (periods of years) tests need to be carried out to gain a mechanistic chemical understanding of the leaching behavior as distinct from the raw numbers in the prescribed leach tests (see Section 5.19.7). An important facet of leaching and long-term aqueous durability is the existence or otherwise of natural analogs of the phases making up HLW waste forms, because if the natural mineral can be found to exist in a wet environment, knowledge of the local geology can give information on the time of exposure to water, and measurements of trace quantities of natural radionuclides in the mineral (such as U, Th, K, Rb) and their daughter products can allow the age of the mineral to be determined.

491

In favorable circumstances, it can be determined that mineral analog phases can last up to millions of years in wet environments, just what is needed for the manmade phases for the sequestration of HLW. Thus, again, there is a powerful argument to use waste forms based on natural analog minerals that have demonstrated their survival over geological time frames. The well-known Oklo phenomenon in the African country of Gabon is worthy of mention in this context. Roughly 2000 Ma, enough U had aggregated in a geological formation there to form an intermittently self-sustaining nuclear reactor in the presence of water to moderate the neutron flux. At that time, the natural 235 U content of U was around 6%, as distinct from 0.71% today. The principal evidence for nuclear activity is a low abundance of 235U in the residual U-bearing material. It is reassuring that the movement of the fission products having half-lives long enough to be still active has been shown to be only a few meters at most away from the residual ore body. A recent description of the disposition of fission products at Oklo has been given by Hidaka.28 Single-phase ceramics have been widely advocated for both single radioactive elements formed by partitioning of reprocessing wastes or even for the entire complement of waste elements. NZP structures have been widely studied and/or advocated for the full range of fission products and actinides.29–31 Monazite, apatite, and zircon have been studied to immobilize actinides, while pollucite10 and CaAlSi5O1232,33 have been investigated for Cs immobilization. However ‘single phase waste forms’ lack chemical flexibility. An exact match of waste and precursor stoichiometries in single-phase multication hosts, such as those mentioned earlier, is industrially unrealistic when dealing with radioactive waste materials. What is needed is an ‘extra’ minor durable phase whose abundance may vary as the waste/precursor ratio varies while still maintaining the same qualitative phase assemblage-as in the synroc-type ceramics. As mentioned previously, the sintered supercalcines10 consisted of apatite and monazite phosphates, powellites, feldspar, pollucite, etc. But there are difficulties of diluting it with materials such as alumina, silicate, or phosphate to deal with radiogenic heat production, apart from the very inelegant approach of using cold fission products as diluents.34 Also, volatility losses during sintering would be severe, although these could be minimized by heating in neutral or reducing atmospheres, and using HIPing. The Rockwell Science Center (CA, USA) realized this latter factor as early as 1981, and they put

492

Ceramic Waste Forms

forward HIPing as the preferred consolidation method for their ceramics35 directed at the tank waste-type HLWs at the Savannah River Laboratory. The ceramics based on alumina tailoring contained magnetoplumbite [Ca(Al,Fe)12O19], UO2, spinel [Mg (Al,Fe)2O4], nepheline (NaAlSiO4), and corundum, with the former phase being seen as a near-universal solvent for fission products other than gaseous species. A titanate-based ceramic was broadly similar to the synroc-D ceramic (the following section) and contained zirconolite, nepheline, spinel, magnetoplumbite, perovskite, murataite (a complex fluorite-based phase), and glass. The waste loadings were around 60 wt%, and HIPing was carried out at 1040  C/60 MPa. There were no problems in these materials with radiogenic heat because the waste was quite dilute in fission products.

5.19.5 Titanate Ceramics Ringwood et al.,12–14 devised ceramics containing phases based on durable natural titanate minerals. These ceramics were called ‘synroc’ (synthetic rock). To deal with Purex-type waste, these theoretically dense materials are made by first mixing inactive precursors of Al, Ba, Ca, Ti, and Zr oxides with liquid (simulated) HLW, drying, and calcining in a H2/N2 atmosphere for 1 h at 750  C. The calcine was then mixed with 2 wt% of powdered Ti metal for redox control and then subjected to uniaxial graphite die hot-pressing or HIPing at 1100  C. The precursor composition and the titanate phases in the early synroc-C titanate ceramic designed for reprocessed commercial power reactor wastes are given in Table 4. Since 1984, rather than using oxides, a slurry mixture of Ba and Ca hydroxides and

transesterified Al, Ti, and Zr alkoxides has been used as the precursor. This provides better solid-state reactivity than the corresponding powdered metal oxides and hydroxides. The principal advantage of this synroc-C ceramic was that the waste ions were dilutely incorporated in durable titanate mineral phases that were considerably more insoluble in water than the silicates and phosphates, and the like used in supercalcine. The waste loading could be varied between zero and 35 wt% using the same inert additive chemistry without substantially changing the basic zirconolite þ perovskite þ hollandite þ rutile phase assemblage, although of course the percentages of the different phases varied somewhat.36 This flexibility is seen as a large advantage. There were minor alumina-rich phases in the more dilute formulations. The grain size is on the order of 1 mm (Figure 4) to optimize mechanical properties and prevent subsequent radiation-induced microcracking (see Section 5.19.8). For comparison, the grain size of a synroc-C sample prepared by sintering at 1300  C is somewhat larger (Figure 5). The alloy phases derive from ions that form metals under the reducing conditions prevailing during hot pressing. These ions are Mo, Tc, Pd, Rh, and Pd, plus any corrosion products from stainless steel. The leach rates at 90  C in water from synroc-C of the most soluble elements, alkalis and alkaline earths, are typically <0.1 g m2 day1 for the first few days, and they decrease asymptotically to values of 105 g m2 day1 after 2000 days (Ringwood et al.14 and see Figure 6). Leach rates of other elements are much

Table 4 Composition and mineralogy of synroc-C (20 wt% reprocessing waste loading) Phase

wt%

Radionuclides in lattice

Hollandite, BaAl2Ti5O14 Zirconolite, CaZrTi2O7 Perovskite, CaTiO3 Ti oxides, mostly TiO2 Alloy phases

30

Cs, Rb

30

RE, Zr, An

20 15

Sr, RE, An None

5

RE: rare earth; An: actinides.

Tc, Pd, Ru, Rh, Mo, Ag, Cd, Se, Te

6864

1.5 kV

X1000

10 µm WD37

Figure 4 Backscattered electron micrograph of Synroc-C. Large feature at bottom right is a partly oxidized lump of Ti metal that was added for redox control. Bright spots are metallic alloys and the remaining micron-sized features are the ceramic phases, rutile, hollandite, perovskite, and zirconolite.

Ceramic Waste Forms

493

Table 5 Seven-day MCC-1 leach results for different elements in synroc-C

P

R

Element

Leach rate (g m2 day1)

Element

Leach rate (g m2 day1)

Mo Cs Tc Ru Sr Ca

0.4 0.1 0.05 0.03 0.02 0.02

Al Zr Ti RE An

<0.4 8  104 2  104 104 – 103 2  105 – 5  104

RE: rare earth; An: actinides.

9636

1.5 kV

X3000

10 µm

WD 8

Figure 5 Synroc-C prepared by pressureless sintering at 1300  C. The black phase is rutile (R). The dark-gray phase is perovskite (P), the white spots are the metallic alloys, and the matrix is a mixture of Ba-hollandite and zirconolite of similar contrast.

Differential leach rate (g m–2 day–1)

1.0E+00

1.0E−01

1.0E−02 Cs 1.0E−03

Ba

Sr

Mass 1.0E−04 0

10

20

30

40 50 60 Time (days)

70

80

90

Figure 6 Leaching of synroc-C in seven-day Materials Characterization Center-1 test.

lower (see Table 5). Leach rates of 105 g m2 day1 correspond to a corrosion rate of ≲1 nm day1. In parallel, a HIPed synroc-D formulation having comparable performance to synroc-C was put forward in conjunction with the Lawrence Livermore National Laboratory (LLNL) in the United States13,37 to deal with the SRL defense waste, and this synroc derivative was based on zirconolite, perovskite, spinel, and nepheline. In the 1980s, the inactive Synroc production process was scaled-up via the Synroc Demonstration

Plant at ANSTO (then the Australian Atomic Energy Commission) to produce 50 kg monoliths containing 20 wt% of simulated Purex HLW (synroc-C), with leaching and microstructural properties as good as those of gram-sized laboratory samples. In the early 1990s, the synroc ceramics were tailored toward the study of zirconolite-rich materials for immobilization of actinide-rich wastes such as Pu or partitioned transuranic elements. The initial work during 1991–1994 was directed at the latter application in conjunction with the Japanese Atomic Energy Research Institute. There was a strong focus on radiation damage via the incorporation38 of the a-emitter 244 Cm (18 year half-life), as had been done with synroc-C and a Na-doped variant thereof.39 Perovskite was also studied for comparison. The work on surplus Pu immobilization, with LLNL as the lead laboratory for the USDOE, moved from zirconolite- to pyrochlore-rich ceramics during 1994–1997. This was because of solid solution limits in the first instance when the target of the work changed from immobilization of 10 wt% Pu (impure) alone to the additional inclusion of 20 wt% U. The estimated time for amorphization of these ceramics to be complete is on the order of 1000 years and the resultant volume expansion would be around 6%.40 This expansion in polycrystalline samples doped with 238Pu, which became X-ray amorphous after around 2 years’ storage at ambient temperatures, did not produce microcracking, and no significant radiation-enhanced aqueous dissolution rates were observed with the crystalline ! amorphous transition. In addition, 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. Nearfield aggregation of Pu due to leaching was shown to be not a problem from the criticality aspect either, because the measured leach rates of Pu were spanned by those of the neutron absorbers5: hence, any leached

494

Ceramic Waste Forms

1 µm 10 wt% HLW, cooled 20 ⬚C min–1 Figure 7 Pellets of Pu-bearing ‘hockey pucks’ prepared by sintering.

Pu would be accompanied by neutron absorbers, which in turn would inhibit criticality problems. The final baseline (no impurities) version4 of the pyrochlore-rich ceramics chosen by the US DOE in 1998 contained 95 wt% of a pyrochlore-structured Ca0.89Gd0.23Hf0.23U0.44Pu0.22Ti2O7 phase plus 5 wt% of rutile-structured Ti0.9Hf0.1O2. The form of the ceramic was to be 76-mm diameter pellets weighing 500 g (Figure 7). Five hundred and sixty such pellets were to be enclosed in a US standard canister of Savannah River DWPF glass to provide a radioactive barrier (gamma field) to prevent diversion. This product was the first crystalline material to be validated in the United States. However, in early 2002, it was decided to remove the disposal option for US/Russian surplus Pu and to proceed only with a MOX fuel option for utilization. This latter option however has not been realized. While it has been realized that substituting Zr for Ti in these ceramics would severely limit the amount of radiation damage sustained by the pyrochlore lattice,41,42 it has to be stressed that this would both increase the fabrication temperature by roughly 300  C and severely restrict the entry of impurities in the target impure Pu into the pyrochlore structure.43 Other synroc derivatives have been devised for immobilization of Tc44,45 and Cs/Sr/Rb/Ba19 formed when reprocessed waste is subjected to partitioning into chemically similar groups.

5.19.6 Glass–Ceramics Glass–ceramics in principle combine the advantages of glasses and ceramics. They can be made by melting, cooling, and reheating at 1000  C to induce

Figure 8 Phase separation shown by scanning electron microscopy in sphene glass–ceramic body after furnace cooling from melt. The continuous phase is Si-rich and the discontinuous phase is a (Ca, Ti, Si)-rich phase that forms sphene on reheating. Etched for 30 s in 1% HF solution.

crystallization, or processed at subsolidus temperatures and slowly cooled to ambient temperatures. Careful design can produce crystalline mineral analog phases chosen for their immobilization qualities together with a durable glass that can provide further tolerance for variations in the waste/precursor ratios and variations of the waste feedstock. Sphene glass–ceramics were developed for 6 years in Canada for HLW arising from a reprocessing option, until it was eventually decided in 1984 to follow the United States and concentrate on spent fuel immobilization. The Canadian glass–ceramics consisted of sphene, CaTiSiO5, in a durable aluminosilicate matrix. The overall composition of the wastefree precursor in mol% was: Na2O (6.6); Al2O3 (5.1); CaO (16.5); TiO2 (14.8); SiO2 (57.0) and considerable variations in this composition were allowable without compromise of the essential properties. The material was produced by melting at 1350  C, cooling to ambient conditions, then reheating for 1 h or so at 950–1050  C to allow the sphene to crystallize within the durable glass phase.46,47 Considerable phase separation occurred during the postcooling step (Figure 8). Loadings of Purex-type HLW were feasible, although additional perovskite and other phases were observed at waste loadings of >10 wt% fission product oxides. Workers at the Hahn-Meitner Institute in Germany studied the properties of borosilicate glasses containing Purex-type HLW and the glasses were devitrified. Different formulations yielded celsian (BaAl2Si2O8), fresnoite (Ba2TiSi2O8), diopside (CaMgSi2O6), or perovskite as major crystalline phases.48,49 The best versions were the materials yielding celsian and these were also studied in the United States. In the United States,

Ceramic Waste Forms

different groups studied the glass–ceramics derived from melting mixtures of natural basalt powder and HLW calcines.50,51 Hanford (WA, US) tank wastes are rich in alkali nitrates and transition metal hydroxides, and a range of glass–ceramics was designed for these.52,53 The 4400 m3 of calcines stored at the INL are rich in alumina, zirconia, and CaF2. While only about 30 wt% of these calcines can be incorporated in glass,54 glass–ceramics studied in the late 1980s and early 1990s and produced by HIPing to immobilize the calcines had waste loadings of around 70 wt%.55 These utilized SiO2-rich frit additives. Subsequently, ANSTO workers in unpublished reports have recently developed separate glass–ceramics for immobilization of the alumina-rich and the zirconia-rich ICPP (Idaho Chemical Processing Plant) calcines. Actinides in various HLWs have been preferentially partitioned toward synroc phases, principally zirconolite, in boroaluminosilicate glass matrices (unpublished work at ANSTO, Loiseau et al., 56 Mahmoudysepehr and Marghussian57). These glass–ceramics have waste loadings of 30–80 wt% and leach rates are often 10–100 times lower than those for standard US EA glass, the baseline glass to pass the product consistency test (PCT) leach test (see next section). These glass– ceramics were prepared by melting, apart from the ANSTO work in which the HIP method described in Section 5.19.10.2 was used.

5.19.7 Aqueous Dissolution Reactivity with water of solids in the first instance depends on the state of aggregation of the solid and clearly the dissolution rate of a solid body will be less than that of a fine powder. The dissolution rate itself can be quantified by measuring the concentration of dissolved species in the water in relation to the original inventory in the solid before the onset of leaching. Dissolution rates can then be expressed as elemental loss by mass per unit surface area (expressed as geometrical or Brunauer–Emmett–Teller (BET) values) per unit time (see Section 5.19.5). These rates, however, can depend critically on (surface area/liquid volume), pH, temperature, presence of salts in the water, etc. Dissolution rates can be further complicated by the presence of colloids and adherence of primarily dissolved species to vessel walls. However, separation of the solids from the liquid, followed by acidification of the liquid, can dissolve species attached to the leach container walls as well as colloids.

495

Colloids themselves can be detected by light scattering measurements. Ultimately, the dissolution rate of a solid in water is controlled by thermodynamics and various software programs are available to describe the dissolution process, although they are usually limited by lack of basic data for some of the ions in the solid. Many laboratories now use apparatus in which the liquid of interest flows over the solid at a given rate, and the rate is not too high to prevent a measurable concentration of dissolved species to accumulate and not too low as to allow the buildup of high concentrations of dissolved materials and consequent complications by solution saturation. In most repositories, the limiting temperature would be designed as 100  C, so measurements on candidate waste forms are experimentally relatively simple. Key regulatory leaching tests are the Materials Characterization Center (MCC)-158 and PCT-B59 protocols (see Chapter 5.18, Waste Glass), which employ polished flat samples and powders, respectively, that are exposed to hot water. The required time of immobilization for real HLW is variously targeted as 104–106 years, and it is worthy of inquiry as to how real-time measurements can be accelerated. This is very difficult as attempts to accelerate leaching by using higher temperatures or more aggressive solutions are easily compromised because the thermodynamics of the solid–liquid interaction can be grossly affected by such means. In practice, short-term (a few days) leach rates of 1 g m2 day1, normalized to take account of the fractional elemental extraction, rather than the absolute quantities, are considered as satisfactory and these rates correspond to  0.1 mm year1, further noting that the leach rates of solids tend to decrease with increasing leaching time even at high degrees of dilution. This is generally attributed to the presence of ‘active surface sites’ on the cut or polished prepared surface of a candidate solid.

5.19.8 Radiation Damage The radionuclides to be immobilized in an HLW waste form include a, b, and g emitters. The most serious damage to the waste form derives from a-decay, in which the a-particle displaces around 100 atoms in the solid during its 20 mm traverse and more importantly, the heavy a-recoil atom which displaces 1500 atoms over its 20 nm trajectory. b- and g-processes produce ionization damage but very few atomic displacements. These effects

496

Ceramic Waste Forms

have been amply demonstrated in natural minerals that contain small amounts of U and Th and have ages of many millions of years, and they have been essentially reproduced in experiments on synthetic materials doped with a few percent of short-lived actinides (238Pu or 244Cm, which have half lives of 87 and 18 year, respectively). Thus, only radiation damage processes in waste forms hosting significant amounts of actinides, especially Pu and other transuranics, need serious consideration.60 The variety of radiation effects include a crystalline ! amorphous transformation after hundreds or thousands of years, with an associated lattice expansion and an associated decrease of several percent in density (e.g., 16% in zircon, ZrSiO4), the production of lattice defects in solids that do not undergo amorphism, formation of gas bubbles, potential for enhanced leaching, and radiolytic effects in which radiolysis of the water leads to the production of species such as H2O2 that may responsible for enhanced leaching. Careful work at Pacific Northwest National Laboratory (PNNL) by Strachan et al.40 has shown that there is no significant leachability increase in pyrochlore- and zirconolite-based ceramics (see Section 5.19.5) as radiation damage progressively builds up. Although it has also long been argued that these kinds of radiation effects on glasses are relatively trivial, it needs to be remembered that the baseline leachabilities of glasses tend to be some orders of magnitude higher than those of crystalline waste forms. Another effect is ‘transmutation damage,’ due to the ionic size and valence changes that may accompany a- or b-emission. Particular examples are Csþ ! Ba2þ and Sr2þ ! Y3þ ! Zr4þ , where the ionic size decreases are 20 and 30%, respectively, for the full decay schemes. These effects have not been studied in much detail because of the intense radioactivity associated with waste forms containing several percentages of the parent isotopes, but sympathetic valence changes in the matrix ions, for example, Csþ þ Ti4þ ! Ba2þ þ Ti3þ and/or the production of hole centers can help to mitigate the charge changes in these decay series.61 Although for ions with half-lives of several years the radioactivity is extremely intense (terabecquerels per gram) and so severely limits available experimental data for these effects, transmutation effects are amenable to study using current atomistic modeling codes and predictions may soon be made by these techniques.

In polycrystalline ceramics, there are two significant potential effects that can lead to microcracking.62 Microcracking can lead to greatly increased leach rates because of the increase in surface area available for water to contact. First, if the actinide-bearing phase is anisotropic, stresses are set up because of unequal lattice dilatations from radiation damage, and secondly, different actinide concentrations in the different phases lead to unequal lattice dilatations between the different phases. However, these effects can be minimized by minimizing the grain size, and this will be discussed in the processing section later.

5.19.9 Thermodynamic Stability of Multication Oxides The intrinsic thermodynamic stability of crystalline waste forms over glasses has long been discussed. For multication oxides, one question is whether the multication oxide has a lower free energy than the component oxides. If not, the multication oxide is unstable with respect to the component oxides. Another question is whether a multication oxide is stable with respect to a simpler multication oxide. The fundamental experimental method is to measure the heat of dissolution of the oxides in question in molten salts such as Pb borates or molybdates at temperatures of around 700  C, and studies of many candidate ceramic waste form phases have been made by Navrotsky’s group in the United States.63 For pyrochlore-type phases, it has been found that several are unstable with respect to forming perovskite phases, but stable with respect to decomposition into simple oxides, whereas both Zr- and Hf-zirconolites are stable with respect to both decomposition modes.64

5.19.10 Processing of Ceramics and Glass–Ceramics The principal options for the production of refractory ceramic or glass–ceramic materials are sintering, uniaxial hot pressing, HIPing, and melting. As mentioned earlier, it is clearly advantageous if these materials are fully dense or at least devoid of open porosity to prevent ingress of water into the interior of the material. All wastes are calcined to remove organics, nitrates, water, carbonates, hydroxides, etc. If relatively small amounts of material are to be dealt with

Ceramic Waste Forms

(say a few tens of tons only), it might be advantageous to mix the waste with precursors before calcination. Though sintering is the baseline method in industry for making (inactive) ceramics, making dense ceramics via sintering is not a trivial task. Metal oxide phases, preferably with multiple cation sites to allow substitution of a variety of fission products, actinides, and process chemicals form the main constituents of ceramic waste forms, and it is necessary to achieve good mixing at an early stage of processing. Best mixing would utilize water-soluble liquid precursors such as metal nitrates to achieve atomic-scale mixing for highquality homogeneous ceramics, but this is easily seen to probably constitute overkill because the footprint of such a plant would be much larger, there are more process steps and nitric acid–based gases evolved which need to be dealt with as a separate (low-level) waste materials. Moreover, if Pu-bearing or enriched U-bearing materials are being immobilized, the use of water presents a criticality risk. So, it is not surprising that the synroc derivative (see above) for surplus US/Russian impure Pu utilized dry feeds that were attrition milled to achieve mixing and reactivity with the dry precursors. MOX fuel is also made by dry powder milling and sintering. These dry operations are most useful if radioactive volatile losses upon sintering are very small, as in the two cases just mentioned, but if volatile losses are potentially significant, HIPing of waste form ceramics has advantages (see Section 5.19.10.2).

5.19.10.2

497

Hot Isostatic Pressing

In HIPing of ceramics or glass–ceramics, the reactive calcined waste form (waste þ additives) material is first packed by vibratory means inside a relatively thin-walled metal can. This is then evacuated and heated to 300–600  C for several hours to remove adsorbed gases, sealed, and then consolidated to full density by compressing it with several tens or even hundreds of MPa of argon gas during the heating cycle. The use of a suitable metal container, which may be stainless or mild steel, nickel, or other metals, can help to achieve the desired redox conditions, minimize any potentially deleterious reaction between the waste form and the container, 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 final consolidation (roughly the same as those used for vitrification, that is, 1000– 1400  C). Figure 9 shows the steps involved in HIPing. Figure 10 shows stainless steel HIP cans before and after consolidation, and Figure 11 shows a diametrically cut section. The process inherently has a batch approach but cans containing more than 100 kg are feasible, with a processing time of 10 h. Work to shorten this time

Radioactive wastes

5.19.10.1 Hot Uniaxial Pressing Hot Uniaxial Pressing (HUP) is a batch process. An inductively heated graphite die can be used and this imposes reducing conditions on the sample, even if the sample is contained in a collapsible metal container or can within the bore of the graphite die. Only 20 MPa of pressure can be exerted with graphite, but very high temperatures (>2000  C) are possible. However, for waste forms, it is usual to keep the temperature to a maximum of around 1400  C to minimize volatile losses. The use of alumina dies allows the use of more pressure and oxidizing conditions, but it is more difficult to extricate the sample, especially when it is radioactive. Of course, in principle, the hot-pressed sample may be left in the graphite or alumina die, which then would constitute an expensive transport container insert. However, HIPing has basically superseded HUP as a processing option for oxide-based ceramics.

Additives

Mixer/drier

Calcination (if required)

Off gas treatment

Fill and seal HIP canister

Preheat

HIP

Canister disposal

Figure 9 Flowsheet for hot isostatic pressing.

498

Ceramic Waste Forms

Figure 10 Stainless steel cans before and after hot isostatic pressing.

submarines and has been validated at INL as a credible (and advantageous) method of consolidating radioactive ceramic waste forms. Moreover, the method is widely used in industry for preparing inactive ceramics. A large advantage is also the relatively small footprint of HIP equipment, arising in the first instance because of the absence of off-gas in the hotconsolidation step. Moreover, the main part of the HIP equipment can be located outside the hot cell so that the HIP does not experience significant radioactive contamination and therefore require disposal at the end of the waste treatment campaign. Also, the HIP process can be used for encapsulation in metal for some wastes. Such examples that have been demonstrated inactively are Sn encapsulation of 129I sorbed on zeolites18 and unpublished ANSTO work on Cu encapsulation of spent fuel pellets and zircalloy liners. For radioactive ceramic waste forms, a prime advantage is to achieve theoretical density with minimum temperature and therefore minimum grain size, thereby adding to the overall strength and reducing the potential of microcracking via radiation damage when the waste form contains a substantial amount of a-emitting waste actinides. In addition, it has been shown for several types of ceramic waste forms that HIP can/ceramic interactions are not deleterious.65–67 5.19.10.3

Figure 11 Diametrically cut can that was hot isostatically pressed.

using hot calcined powders instead of allowing them to cool to ambient temperatures is under way, and throughputs of tonnes per day are targeted. Figure 10 shows that the consolidated can has a basic cylindrical shape, that is part of the can design allowing it to occupy a minimum of space in a cylindrical US transport container. While the relationship of the shapes of the can before and after HIPing is quite complex, the basic variable is the ratio of the densities of the final ceramic and the calcined powder. The dumbbell shape of the can gives quite a deal of flexibility, but it is advantageous to maximize the density of the calcined powder to avoid undue rippling and substantial deviations from cylindrical geometry of the HIPed can. The HIP process, invented by the Battelle company in the United States in the 1950s, has been used since the 1960s in preparing nuclear fuel for

Melting

Joule melters employ large refractory ceramic baths (several square meters in area and 1 m or so deep) containing ceramic electrodes to directly heat mixtures of glass frit and nuclear waste to a molten state. The melt is then poured into steel transport canisters in which it cools slowly to yield a borosilicate glass. Such melters are only viable in the longer term if temperatures do not exceed 1150  C. For higher temperatures, cold crucible melters are necessary. Cold crucible melting utilizes inductive coupling between a water-cooled high-frequency coil and conductive waste form.68 This coupling is inclined to be low when heating is commenced, so metal or graphite often needs to be added to the ceramic charge. This oxidizes and the metal oxide combines with the charge when high temperatures are reached, while the graphite is lost as CO2 under these conditions. While some ceramics such as synroc-C can readily be melted at temperatures below 1500  C,69,70 there are serious questions of fission product volatility (captured and recycled), especially if reducing conditions are not maintained. Moreover, on cooling from the melt

Ceramic Waste Forms

at rather low cooling rates governed by the large size of canisters into which the melts would be poured (pouring would not be easy if melts were not produced in an air atmosphere), crystallization of the waste form would allow the formation of quite large grains (Figure 12). This factor would adversely affect the mechanical strength and the response to a-radiation if the wastes contained significant actinide inventories.

attendant fission product volatility can be avoided. The composition of ordinary Portland cement (OPC) is given approximately in Table 6. The curing of cement after mixing it with water and aggregate is complex and takes place over months and years; the phases in the dry cement clinker gradually transform to hydrated phases, with a CaO–SiO2–H2O (C–S–H) tobermorite phase of variable composition as the main contributor to the eventual strength of the material. OPC can be diluted with flyash from coal-fired power plants or ground blast-furnace slags (approximate compositions in Table 7). In addition to utilizing waste materials, the well-known alkali-aggregate deleterious reaction can be minimized. The behavior of such cements in immobilizing LLWs and ILWs has been extensively reviewed.72–74 Work on cements processed under steam pressure conditions at 100–400  C (FUETAP) at the end of the 1970s and early 1980s75 indicated that good compressive strengths and thermal conductivities were achievable, and the derived leach rate was in the zone of 1 g m2 day1 or better for most species, especially if Cs had been sorbed on to zeolite and then incorporated in the cement, apparently showing that cements were intrinsically leach resistant, nearly on a par with borosilicate glasses. However, cements in which the Cs is not presorbed on to zeolites fail the standard leach tests described earlier by factors in the range of 10–100 for ions that do not form insoluble hydroxides – such as Cs (especially in MCC-1 tests in which the sample geometrical surface area to leach liquid volume is low) – and they are not these days seen as serious candidates for HLW immobilization. Also, the presence of water even in a bound state is problematical for radiolytic hydrogen buildup. However, cementitious material is currently seen generally as still having strong potential for LLW and ILW immobilization.

5.19.11 Cements and Geopolymers Cement is at first sight a particularly attractive means of HLW consolidation since high temperatures and

15277

23 µm Figure 12 Microstructure of synroc formulation designed71 for (Al,U)-rich high-level waste and processed by cold crucible melting. Dark areas: spinel, MgAl2O4; Light gray: pyrochlore-structured (Ca,U) titanate phase; darker gray areas: hollandite and rutile.

Table 6

499

Approximate cement and geopolymer compositions (excluding hydrous material and carbonate)

Composition (wt%)

CaO

SiO2

Al2O3

Fe2O3

Na2O

MgO

SO3

OPC Geopolymer

60 5

20 50

5 25

5 5

2 15

5

3

Table 7

Flyash Slag

Approximate blast-furnace slag and flyash compositions CaO

MgO

Al2O3

SiO2

Fe2O3

C

Na2O

K2O

5 40

1 10

25 10

50 40

10

5

3

2

500

Ceramic Waste Forms

Geopolymers are a class of cementitious materials that can be described as alkali-activated cements.76–79 They are made by reacting at ambient temperatures aluminosilicates such as metakaolin, fly ash, or ground blast-furnace slags with alkaline solutions, usually strong NaOH solutions. The baseline stoichiometries (Table 6) for the materials participating in the reaction are typically Na/Al ¼ 1 and Si/Al ¼ 2 on a molar basis, maximizing the strength80 and a minimum amount of water (H2O/Na  7 on a molar basis) is used to assure approximately a maximum amount of reaction. The properties are relatively insensitive to variations in the molar ratios at the level of 10 or 20%. The aluminosilicates partly dissolve in the solutions and polymerize and solidify. Curing is carried out at 40–90  C. Extensive studies by solid-state nuclear magnetic resonance have been carried out over the years together with porosity studies, so it is now accepted81 that geopolymers consist of nanoporous aluminosilicate networks, with water in the pores, although micro- and macroporosity is also present. Further evidence for nanoporosity has been gleaned from transmission electron microscopy and mercury porosimetry82 and positron annihilation lifetime measurements.83 Samples that pass the PCT aqueous dissolution test84 (but which have high MCC-1 leach rates of alkalis) can be fabricated by this technique. Systematic studies of the aqueous leaching behavior of geopolymers have not yet been carried out, but measurements of time dependence (1–90 days) and surface area/volume ratios suggest that at ordinary temperatures (90  C) the principal leaching mechanism derives from exchange of the pore water with the leaching solution, rather than attack of the aluminosilicate framework. As expected from the strength measurements,80 metakaolinite-based geopolymers having Na/Al  1 and Si/Al  2 molar ratios have maximum aqueous durability.85,86 Further measurements in progress at ANSTO are looking at the effects of temperature over the range of 20–90  C, pH in the range of 2–12, and the effect of bicarbonate and chloride ions in the leaching solution. Geopolymers have better fire and acid resistance than standard OPC. Moreover, geopolymers are serious candidates for ILW, especially as they can be dewatered by heating to 300–400  C without significant effects on their mechanical and chemical properties87 as long as the thermal ramp rate is kept fairly low to minimize structural disruption from the egress of the water. This is to be contrasted with OPC in which the C–S–H strength-building phase is decomposed at

temperatures above 200  C, thereby severely impacting the mechanical properties. Geopolymers have been used in Slovenia and Kazhakstan to immobilize large amounts of ILWs.88 Magnesium potassium phosphate (MPP) ceramics based on MgKPO46H2O have been developed at the Argonne National Laboratory since the early 1990s.89,90 Inactive versions of these materials have also been used in fertilizers for agricultural purposes.91 These materials can in some sense be regarded as cementitious or low-temperature ceramics. The distinction is perhaps academic as phosphate cements are well known. MPPs are prepared by mixing calcined MgO with strong solutions of KH2PO4 and allowing the following reaction to take place: MgO þ KH2 PO4 þ 5H2 O ! MgKPO4  6H2 O While it has recently been suggested that MPPs are appropriate for HLW immobilization,92 the leach rates in that work were based on BET surface areas. Perhaps more importantly, the hydrous nature of MPPs gives rise to a radiolytic H2 hazard, and unpublished experimental work at ANSTO has shown that MPPs become very weak structurally when heated above 400  C for the purposes of dewatering and/ or attempted conversion to anhydrous ceramics.

5.19.12 Conclusions In spite of more than 40 years of work, the disposition of high-level nuclear fuel wastes around the world in future is still subject to many uncertainties, especially with Yucca Mountain as an HLW repository in the United States being currently abandoned. Apart from political and NIMBY arguments, much of the scientific debate surrounds the question of how to validate physical models that lead to the calculated maximum radiation dose to persons living close to the repository, and more particularly how to convince a lay audience that the very complex calculations, including the uncertainties, are meaningful. However, the waste form is a key containment barrier because it can be subjected to rigorous experimental study, and optimization of its behavior can be studied directly at least over a few years. In this respect, more analog studies of natural minerals are needed, where although the water/thermal history of the analog mineral itself may be hard to derive, the history may well be derivable from the surrounding minerals in the rock formation. It seems clear that

Ceramic Waste Forms

Table 8 Radioactive HLW particularly amenable to ceramification over vitrification Waste

Difficulty for vitrification

Contaminated metalsa High-Al and high-Zr wastes (ICPP, for example) Tc and Cs Actinides

Incompatible with glass Low solubility in silicate glass Volatile losses Low solid solubility of lower actinide valence states

a

Can be encapsulated in ceramics using HIPing.

5.

6. 7. 8. 9.

waste form development for the large spectrum of chemically distinct HLWs already in existence for many years, plus those yet to be generated by ongoing and future nuclear power programs, will continue. The twin foci of this continuance are simply (a) increased waste loading, especially in cases where radiogenic heating is not serious, to ease the amount of space required to contain the waste forms in repositories and (b) cost savings via development of technologies that maximize waste form throughputs and minimize plant footprints and radioactive offgas emissions. It is certain that HIPing will play an important role in achieving these improvements, particularly for wastes which are problematic for vitrification (Table 8) and for which there will be a strong preference for ceramics and glass–ceramics. Cements and particularly geopolymers remain as potentially viable for ILW. (See also Chapter 5.22, Minerals and Natural Analogues; Chapter 5.18, Waste Glass and Chapter 1.05, Radiation-Induced Effects on Material Properties of Ceramics (Mechanical and Dimensional)).

10. 11.

12.

13.

14.

15. 16. 17.

18. 19.

Acknowledgements

20.

The author wishes to acknowledge numerous colleagues at ANSTO and around the world for many discussions and contributions over many years.

21.

References

23.

22.

24. 1. 2.

3. 4.

Ewing, R. C. MRS Bull. 2008, 33, 338–341. USDOE. The Evaluation and Selection of Candidate High-Level Waste Forms; US Department of Energy Report DOE/TIC 11611; Savannah River Operations Office: Aiken, SC, 1982. Hench, L. L.; Clark, D. E.; Campbell, J. Nucl. Chem. Waste Manag. 1984, 5, 149–173. Cochran, S. G.; Dunlop, W. H.; Edmunds, T. A.; MacLean, L. M.; Gould, T. H. Fissile Material Disposition

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