Waste Forms Technology and Performance: Final Report (2011)

3

Waste Forms

The final charge of the statement of task for this study (see Box 2.1 in Chapter 2) calls for the identification and description of “potential new waste forms that may offer enhanced performance or lead to more efficient production.” This chapter, which is written primarily for technical audiences, addresses this charge by providing a brief review of waste form materials and an assessment of their potential applicability to Department of Energy, Office of Environmental Management (DOE-EM) waste streams. Possible applications are also highlighted in Chapter 1.

A voluminous technical literature on waste form materials has developed over the past six decades. A comprehensive review of this literature is well beyond the scope of this study. However, the committee has included key historical and review article references in this chapter for interested readers.

3.1 WASTE FORM DEVELOPMENT

The concept of immobilizing radioactive waste in either vitreous or crystalline materials is more than 50 years old. In 1953, Hatch (1953) of Brookhaven National Laboratory introduced the concept of immobilizing radioactive elements in an assemblage of mineral phases. The first borosilicate glass formulations were developed in the United States between 1956 and 1957 by Goldman and others at the Massachusetts Institute of Technology (Eliassen and Goldman, 1959; Goldman et al., 1958; Mawson, 1965). These researchers examined calcium-aluminosilicate porcelain glazes

to which boron oxide (B₂O₃) had been added to achieve a pourable glass and minimize radionuclide volatilization. The most promising vitreous systems for future development were determined to be borosilicate based, e.g., CaO-Al₂O₃-B₂O₃-SiO₂ and Na₂O-CaO-Al₂O₃-B₂O₃-SiO₂.

In 1970, the singular requirement for a waste form from the Office of Nuclear Waste Isolation¹ (ONWI) was that it be a stable solid (DOE, 1981; Walton et al., 1983). By the mid-1970s, innovative proposals for producing stable solid waste forms were being offered—for example, supercalcine ceramics by Rustum Roy and colleagues at Pennsylvania State University (McCarthy, 1977; Roy, 1975, 1977, 1979); alumina-based tailored ceramics by Rockwell International Science Center (Jantzen et al., 1982b; Morgan et al., 1981); and titania-based SYNthetic ROCk (SYNROC) by Ted Ringwood and colleagues at the Australian National University and the Australian Nuclear Science and Technology Organisation (Reeve et al., 1984; Ringwood, 1978, 1985; Ringwood et al., 1978). The first systematic compilations of potential crystalline waste form phases were also made at this time (Haaker and Ewing, 1981).

There were extensive research and development (R&D) programs on nuclear waste forms during the late 1970s and early 1980s, resulting in the examination of a wide variety of single-phase and polyphase ceramics. By this time “low leachability” had become the main criterion for waste form comparisons (DOE, 1981; Walton et al., 1983), and such comparisons between crystalline ceramics and glass generated considerable controversy (Kerr, 1979a,b).

Beginning in 1978, there was intense study of alternative waste forms that culminated in a review (Garmon, 1981) that recommended borosilicate glass for immobilizing high-level radioactive waste (HLW) at the Savannah River Site (SRS) in South Carolina and West Valley in New York and also identified SYNROC/tailored ceramics as promising alternatives (Hench et al., 1981). Glass was considered to be a more proven technology, and there were questions about the maturity of production technologies for ceramic waste forms. Nevertheless, Hench et al. (1981) made a strong recommendation for continued research and development for ceramic waste forms, including SYNROC and titanate- and alumina-based ceramics. These alternative waste forms were later determined to be difficult to process, more costly to implement, and not as flexible for accommodating variations in waste composition as borosilicate glass (De et al., 1976; Dunson et al., 1982; Lutze et al., 1979; McCarthy, 1973; McCarthy and Davidson, 1975; Morgan et al., 1981; Ringwood et al., 1981; Schoebel, 1975), even though

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¹ The Office of Nuclear Waste Isolation was located at the Battelle Memorial Institute. It conducted research and published technical reports on technical aspects of nuclear waste isolation.

many were found to have superior product quality (Hench et al., 1981; Walton et al., 1983).

High-temperature processing of these alternative waste forms frequently resulted in the formation of an intergranular glass phase, especially when alkali-containing wastes were processed. This intergranular glass limited product stability and durability because radionuclides such as cesium-137 and strontium-90, which were frequently incorporated into the intergranular glass phases (Buykx et al., 1988; Clarke, 1981; Cooper et al., 1986; Zhang and Carter, 2010), were determined to leach at the same rates as those from glass waste forms (Jantzen et al., 1982a). Because little was understood at the time about the degradation mechanism of a single-phase glass versus glass-ceramic materials (i.e., materials that contain both glass and crystalline phases), borosilicate glasses were selected for continued development over the alternative waste forms (Walton et al., 1983).

Research activity on alternative waste forms was severely curtailed as a result of the 1981 decision in the United States to immobilize defense HLW in borosilicate glass and the subsequent construction of the Defense Waste Processing Facility (DWPF) at SRS and the West Valley Demonstration Project (WVDP) at West Valley. The R&D effort on nuclear waste forms during this period has been summarized by Lutze and Ewing (1988).

More recently, there has been a resurgence of interest in crystalline waste forms because of the need to develop durable materials for the stabilization and disposal of actinides such as plutonium from defense and civilian programs (Burakov et al., 2010; Ewing, 1999; Ewing et al., 1995b; Oversby et al., 1997). There has been additional R&D work on minerals and their analogues (e.g., apatite, monazite, zirconolite, zircon, and pyrochlore) (Ewing et al., 1995a) and SYNROC formulations (Ryerson and Ebbinghaus, 2000) as well as another down selection between glass and ceramic waste forms (Meyers et al., 1998).

Crystalline waste forms made from clay have been studied almost continuously since 1953 (Hatch, 1953; Lutze et al., 1979). Roy (1981) proposed low-temperature, hydrothermally processed, low-solubility phase assemblages consisting of mineral analogues of mica, apatite, pollucite, sodalite-cancrinite, and nepheline, many of which could be made from reactions between clays (kaolin, bentonite, and illite) and waste. Mineral analogue waste forms made from clays have been recently re-examined for the immobilization of high-sodium, salt supernate HLW at the Hanford Site in Washington; high-sodium recycle streams from tank cleaning at the Idaho National Laboratory (INL), and low-activity waste melter off-gas condensates at Hanford. These mineral analogue waste forms are made using a moderate-temperature (700°C-750°C) thermal pyrolysis treatment (Mason et al., 1999, 2003) (i.e., steam reforming; see Chapter 4) by adding clay to the waste to form feldspathoid mineral analogues (sodalite and

nepheline) or dehyroxylated mica (Jantzen et al., 2008), depending on clay composition.

Stabilization and solidification with cement-based binders has been used to immobilize radioactive wastes since the beginning of the nuclear age. The process has been used to encapsulate solid waste, solidify liquid waste (including tritiated water), stabilize contaminated soils, stabilize tank-heel residues after tanks are emptied, and as low-permeability barriers. Cements have also been used as binders and to encapsulate granular or cracked waste forms. A recent comprehensive review of cement systems for radioactive waste disposal can be found in Pabalan et al. (2009). Long-term cement durability comparisons have been made using ancient cements, geopolymers, and mortars (Jiang and Roy, 1994; Kovach and Murphy, 1995; Krupka and Serene, 1998; Miller et al., 1994; Roy and Langton, 1983, 1984, 1989; Steadman, 1986), some of which may also serve as natural analogues for geopolymer waste forms (Barsoum et al., 2006, but see also Jana, 2007).

Recent reviews of developments in waste form research are provided in the following papers: Caurant et al. (2009); Donald et al. (1997); Ewing (1999, 2001); Ewing et al. (2004); Lee et al. (2006); Lumpkin (2001, 2006); Lutze and Ewing (1988); Ojovan and Lee (2005, 2007); Stefanovsky et al. (2004); Weber et al. (2009); and Yudintsev et al. (2007). The most recent interest has been associated with the desire to create new waste forms as part of advanced nuclear fuel cycles involving recycling of irradiated fuel (Peters and Ewing, 2007). Recent reviews of radiation effects in waste forms can be found in a series of papers by Ewing and others (1995b); Ewing and Weber (2010); and Weber and others (1997, 1998). Reviews of natural analogues that provide long-term data on the durability of glass and crystalline ceramics have been provided in a number of papers, including Allen (1982); Ewing (1979, 1999); Haaker and Ewing (1981); Jantzen and Plodinec (1984); Malow et al. (1984); Morgenstein and Shettel (1993); and Verney-Carron et al. (2010).

3.2 ROLE OF WASTE FORM IN WASTE IMMOBILIZATION

As noted in Chapter 2, the primary role of a waste form is to immobilize radioactive and/or hazardous constituents (hereafter simply referred to as constituents) in stable, solid matrices for storage and eventual disposal. Immobilization can occur through chemical incorporation, encapsulation, or a combination of both processes. Table 3.1 provides a pictorial representation of the different combinations of chemical incorporation and encapsulation for the waste form materials described in Section 3.3.

Encapsulation is achieved by physically surrounding and isolating constituents in a matrix material, which traps waste ions on grain boundaries and in some cases sequesters constituents in hydrated products. Cements,

TABLE 3.1 Key Properties of the Waste Form Materials Described in this Chapter

Waste Form Class (see text)	Retention Mechanism	Graphical Representation	Waste Form Properties	Examples
1. Single-Phase Glasses Constituents are atomically bonded in the glass structure, usually to oxygen that is also bonded to other matrix elements (e.g., Si, Al, B, P) by short-range order (SRO) and medium-range order (MRO).	Chemical incorporation		a. Moderate waste loading b. Good overall durability c. Easy to model constituent release from a single phase	Borosilicate glasses, aluminosilicate glasses, phosphate glasses
2a. Glass-Ceramic Material Constituents are present in the glass matrix, and benign crystals such as spinels (Cr, Ni, and Fe species) ( ). These are allowed crystals to do crystallize not contain radionuclides but may contain hazardous constituents (e.g., Cr, Ni).	Chemical incorporation		a. Higher waste loadings for high Cr, Ni, and Fe wastes b. Good overall durability c. Easy to model constituent release from single-phase glass because there is minimal impact from grain boundary dissolution (has to be determined by experimentation)	Higher waste-loaded borosilicate, alumino-silicate, or phosphate glasses

Waste Form Class (see text)	Retention Mechanism	Graphical Representation	Waste Form Properties	Examples
5, 6, 7, 8. Cements, Geopolymers, Hydroceramics, Ceramicretes Hydrated phases weakly incorporate constituents or retain them by sorption. Encapsulation is by solidification or precipitation of constituents on grain boundaries where non-constituent phases hydrate or crystallize. Example shows Tc sequestered by C-S-H hydrates and sequestered by secondary fly-ash granules.	Encapsulation		a. Low waste loading b. Lower overall durability c. Difficult to model constituent release from multiple phases and hydrated secondary phases d. Easy to process—usually mix and set	Savannah River Site Saltstone

Key:

   Pu   Tc   U   Cs

geopolymers, ceramicrete, and hydroceramics (see Section 3.3) can be used as waste forms or as binders for other waste form materials. Encapsulation is typically used to immobilize low-level or intermediate-level wastes.

Chemical incorporation involves the atomic-scale bonding of radioactive constituents into radiophases of the waste form material, which are durable structures with any combination of short-range order (SRO),² medium-range order (MRO),³ or long- range order (LRO).⁴ Glasses incorporate constituents into their atomic structures by SRO and MRO. Recent experimentation has shown the existence of large, cation-rich clusters in glass. These more highly ordered regions of MRO often have atomic arrangements that approach those of crystals (Box 3.1). Crystalline ceramics incorporate constituents by a combination of SRO, MRO, and LRO. LRO defines the periodic structural units characteristic of crystalline ceramics.

There are two approaches for immobilizing radioactive waste in crystalline materials (Roy, 1975, 1977):

1. Radionuclides can be incorporated into the atomic structure of the phase. Individual radionuclides occupy specific sites in the structure, generally according to atomic size and charge constraints. For complex waste streams, crystalline structures with multiple cation sites are required to accommodate different radionuclides.
2. The radionuclide-bearing radiophases can be encapsulated in another non-radionuclide bearing material to form a composite waste form.⁵ Encapsulating materials, such as TiO₂ or ZrO₂, can have high durability.

3.3 WASTE FORM MATERIALS

A wide range of materials are potentially suitable for immobilizing radioactive waste. For simplicity of discussion, these waste form materials have been grouped into eight classes based on their phase properties:

1. Single-phase (homogeneous) glasses
2. Glass-ceramic materials

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² SRO: radius of influence ~1.6Å-3Å around a central atom, e.g., such as tetrahedral and octahedral structural units.

³ MRO: radius of influence ~3Å-6Å encompasses second- and third-neighbor environments around a central atom. The more highly ordered regions, referred to as clusters or quasicrystals, often have atomic arrangements that approach those of crystals.

⁴ LRO extends beyond third-neighbour environments and gives crystalline ceramic/mineral structures their crystallographic periodicity.

⁵ For example, waste constituents can be chemically incorporated into a crystalline ceramic phase and then encapsulated in another material that provides an additional barrier to release.

BOX FIGURE 2 Proposed polymerization of SRO and MRO in the atomic structure of a French HLW rare earth bearing aluminoborosilicate glass. Na⁺, Ca²⁺, and Nd³⁺ exist in the proposed percolation channels. SOURCE: Caurant et al., (2009).

3. Crystalline ceramics
4. Metals
5. Cements
6. Geopolymers
7. Hydroceramics
8. Ceramicretes

Brief descriptions of these materials are provided in the following sections. More detailed descriptions of these materials are provided in the accompanying tables.

3.3.1 Single-Phase (Homogeneous) Glasses

Glass is an amorphous solid material produced by cooling a material from a molten to a solid state without crystallization. Glass waste forms can have a wide range of compositions, but they are generally classified using

the dominant tetrahedral structural groups forming its structure, similar to the nomenclature for aluminosilicates, borates, and phosphates:

• Borosilicate glasses contain (SiO₄)^–4, (BO₄)^–5, (BO₃)^–3, and some (AlO₄)^–5 structural units.
• Aluminosilicate glasses contain (SiO₄)^–4 and (AlO₄)^–5 units.
• Aluminoborate glasses contain only (BO₄)^–5, (BO₃)^–3, and some (AlO₄)^–5 units.
• Aluminophosphate glasses contain (PO₄)^–3 and (AlO₄)^–5 units.
• Iron phosphate glasses contain (PO₄)^–3 and (FeO₄)^–5 units.

The major properties of glass waste form materials are summarized in Table 3.2.

Glass is being used worldwide to immobilize HLW from reprocessing of spent nuclear fuel and targets. The immobilization process, vitrification, is a continuous process capable of handling large-volume waste streams. This process is a well-demonstrated technology with more than 40 years of industrial experience (see Chapter 4).

Glass has several advantages for HLW immobilization: It can generally accommodate a wide range of waste stream compositions; has adequate-to-good durability in many disposal environments (see Chapters 5 and 8); and has good thermal and mechanical stability properties. In particular, borosilicate glasses melt at temperatures between 1,050°C-1,200°C, which limits the volatility of radionuclides such as technetium-99, cesium-137, and iodine-129. The melts are generally less corrosive than commercial glass melts, such as Pyrex, because of their lower temperatures.

The worldwide production of HLW glass is summarized in Table 3.3. To date, more than 5,000 metric tons of HLW borosilicate glass have been produced at SRS and 500 metric tons have been produced at West Valley, New York. More than 6,700 metric tons of HLW borosilicate glass have been produced in France. The compositions of these glasses, including the incorporated wastes, fall into a common region in the borosilicate glass forming system (Jantzen, 2011; Ramsey, 1989; Wicks et al., 1985). They contain ~60 weight percent or more of glass forming oxides (SiO₂, B₂O₃, ZrO₂, Al₂O₃, P₂O₅, and fission products), >15 weight percent glass modifier oxides (Na₂O, K₂O, Li₂O, CaO, MgO, SrO, and ZnO), and 0-25 weight percent glass intermediate oxides (Cr₂O₃, Fe₂O₃, CuO, NiO, MnO, PbO, TiO₂, and actinides).

3.3.2 Glass-Ceramic Materials

Glass-ceramic materials (GCMs) are materials that contain both crystalline and glass phases (see Table 3.1). Depending on the intended appli-

TABLE 3.2 Single-Phase (Homogeneous) Glass Waste Forms (Type 1 in Table 3.1)

TABLE 3.3 Global Production of HLW Glass at Industrial Scales

^a 1 Tera-Becquerel (TBq) = 10¹² decays per second.

^b 1996-2009.

^c Jantzen et al. (2005).

^d 1996-2002; mission complete.

^e Perez et al. (2001).

^f 1991-2007 at 150 L glass per canister and an assumed glass density of 2.75 g/cm³.

^g Riley et al. (2009).

^h Predicted mission completion is 6582 canisters.

ⁱ Caterine Veyer of AREVA, personnel communication (2010).

^j P.P. Poluektor, personal communication (2010).

^k Acidic waste loadings are comprised of fission products and minor actinides; corrosion products and alkali are not included as for neutralized wastes.

^l 1978-2008

^m 1985-1991

ⁿ 1995-2006

^o Seiichiro Mitsui of Japan Atomic Energy Association (JAEA), personnel communication (2010).

^p 1989-2008

cation, the major component may be a crystalline phase with a vitreous phase acting as a bonding agent. Alternatively, the vitreous phase may be the major component with particles of a crystalline phase dispersed in the vitreous matrix. GCMs can be formed by a number of processes, including melt crystallization (controlled or uncontrolled), multiple heat treatments, or by encapsulation of ceramic material in glass (Lee et al., 2006).

In practice, crystalline materials frequently contain glass remnants along grain boundaries, and unreacted radiophases may persist rather than be completely incorporated into the crystal structures. A virtual continuum exists between 100 percent glass and 100 percent crystalline materials with GCMs falling in between as shown in Figure 3.1.

GCMs offer a useful compromise between glasses and ceramics. They are easier and less expensive to prepare than conventional ceramics but offer higher durability than glasses, provided that the soluble species that can sequester radionuclides (e.g., Na₂SO₄, which can sequester cesium and strontium) (Plodinec and Wiley, 1979) are prevented from crystallizing (Figure 3.1 and Table 3.1).

GCMs offer several potential advantages over glass for use as waste form materials, including increased waste loadings, increased waste form density, and thus smaller disposal volumes. These waste forms can also be used to immobilize glass-immiscible components such as sulphates, chlorides, molybdates, and refractory materials that have very high melting temperatures. They can also be used to immobilize long-lived radionuclides (e.g., actinides) by incorporating them into the more durable crystalline phases; short-lived radionuclides (e.g., many fission products) can be accommodated in the less durable vitreous phase (Lee et al., 2006). Relatively low leach rates (see Chapter 5) have been observed for some glass ceramics, which may make them potential candidates for immobilization of HLW. For example, allowing the formation of crystalline phases that have little to no impact on durability (see Figure 3.1) would be an achievable incremental strategy for increasing HLW loading in glass (and throughput of waste) at SRS and Hanford.

A number of different GCMs have been proposed for the immobilization of HLW; summaries can be found in Donald et al. (1997), Lee et al. (2006), and Stefanovsky et al. (2004). A brief synopsis of GCMs is given in Table 3.4.

FIGURE 3.1 Schematic diagram illustrating the durability and waste loading of waste form materials relative to single-phase (homogeneous) glass. Single-phase glass formulations are shown in the lower left corner of the triangle. If crystals (e.g., spinels) are allowed to form in the glass, then glass-ceramic materials (GCMs) can be produced (left leg of triangle) that have superior waste loading and durability relative to single-phase glass. Fully crystalline materials (upper corner of triangle) are considered to be exceptionally durable waste forms with high waste loadings relative to single-phase glass. However, the incorporation of certain species (e.g., Mo, S, and P) into glass creates non-durable secondary phases (lower right corner or triangle) that may have unacceptable durabilities.

SOURCE: After Ojovan and Lee (2007).

3.3.3 Crystalline Ceramics

Crystalline ceramics are inorganic, non-metallic solids that contain one

TABLE 3.4 Glass Ceramic Materials (Types 2a and 2b in Table 3.1)

TABLE 3.5 Single-Phase Crystalline Ceramic Materials (Type 3a in Table 3.1)

^a In recent usage, zeolite represents the framework topology of the 3D porous crystalline inorganic frameworks (with or without water); it can be applied to the framework of related minerals, their synthetic analogues, and/or non-aluminosilicate porous crystalline 3D frameworks. Structural information from the International Zeolite Association (IZA) website, http://www.iza-structure.org/databases/; see Analcime analogues (ANA).

^b Supercalcines were the high-temperature silicate-based “natural mineral” assemblages proposed for HLW waste stabilization in the United States (1973-1985).

or more crystalline phases.⁶ Single-phase crystalline ceramics (Table 3.5) can be used to immobilize separated radionuclides (e.g., plutonium-239) or more chemically complex waste streams (e.g., HLW). In the latter case, the atomic structure of the ceramic phase must have multiple cation and anion sites that can accommodate the variety of radionuclides present in the waste stream. These materials are potentially attractive for immobilizing long-lived alpha emitting actinides such as plutonium, neptunium, and americium (Burakov et al., 2010). However, some of these materials are susceptible to radiation damage effects associated with alpha decay from actinides; these effects have recently been summarized in Ewing and Weber (2010).

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⁶ However, as noted previously, even materials that are predominantly a single phase will frequently have trace amounts of other phases segregated along grain boundaries or as inclusions of unreacted material.

TABLE 3.6 Multiphase Crystalline Ceramic Materials (Type 3b in Table 3.1)

Multiphase crystalline ceramics (Table 3.6) consist of an assemblage of crystalline phases. Individual phases are selected for the incorporation of specific radionuclides, with the proportions of phases varying depending on the composition of the waste stream. An individual phase can host one or more radionuclides, including solid solutions of radionuclides. However, not all phases will host radionuclides.

The three best known examples of multiphase waste forms are SYNROC, which is based on titanium ceramic phase assemblages; supercalcine ceramics, which are silicate-based ceramic phase assemblages; and tailored ceramics, which are alumina-based ceramic phase assemblages.

3.3.4 Metals

Several different types of metallic materials have been studied as potential waste forms. Like crystalline ceramics, metal waste forms can consist of single or multiple phase assemblages, and the waste form itself can be granular or monolithic. Metal waste forms can be fabricated sintering or casting. Each of these techniques has drawbacks; in particular, it can be

difficult to find metal compositions and processes that effectively wet and encapsulate dispersed phases or fines.

Metal waste forms composed predominantly of lead and lead-based alloys are being used in Russia (and have been used in Belarus and Ukraine) to immobilize radioactive sources (Arustamov et al., 1999; Ojovan et al., 2004). Metal waste forms are being used by DOE-EM to immobilize metallic fuel waste, including fuel hulls, at INL. The waste is solidified by melting at 1,560°C into a uniform, homogeneous waste form. Typically, zirconium is added to the waste, which lowers its melting point and generates enough Fe₂Zr intermetallic phase to accommodate the transuranic elements (Ebert, 2005). As a result, waste loading is typically above 90 percent (Marsden and Westphal, 2006).

Metallic composite waste forms produced by hot isostatic pressing (see Chapter 4) are under study by the Australian Nuclear Science and Technology Organisation (ANSTO) for applications in the United Kingdom (Begg and Smith, written communication). Metal encapsulation will be used to immobilize debris waste streams (e.g., cermets, silicon carbide, graphite, broken fuel pins, fuel hulls) that are not economically feasible to process by other means. The encapsulant material provides a second level of protection to the release of radionuclides.

3.3.5 Cements

Cements are inorganic materials that set and harden as a result of hydration reactions. Cements microencapsulate wastes (Table 3.1), although there is recent evidence that during hydration three binding mechanisms can also occur between the cement and metal ions in the waste (Cocke and Molla, 1993; Cougar et al., 1996; Glasser, 1997):

• Precipitation of metal ions into the alkaline matrix as an oxide, mixed oxide, or as another discrete solid phase.
• Adsorption or (co-)precipitation of metal ions onto the surface of cement minerals.
• Incorporation of metal ions into hydrated cement minerals as they crystallize.

These processes are not mutually exclusive.

The most common type of cement, Portland cement, consists of calcium silicates, other aluminum and iron containing phases, and additives such as gypsum to control set time. Portland cement, when mixed with water and aggregates, hydrates to form concrete. Grout, a mixture of Portland cement and various sand mixtures, is commonly used to encapsulate radioactive

wastes. DOE-EM is filling emptied HLW tanks at SRS and INL with grout to immobilize waste heels and provide structural support.

Cements are used to immobilize waste having relatively low levels of radioactivity (i.e., low- or intermediate-level radioactive wastes). Higher-activity wastes can result in radiolysis and production of hydrogen gas from the breakdown of water or hydroxyl groups in the cement (Bibler, 1980). To minimize radiolysis, the use of concrete formed under elevated temperature and pressure (FUETAP), which reduces the amount of entrapped water to about 2 percent, was considered for the immobilization of HLW (McDaniel and Delzer, 1988; Moore, 1981). This idea was eventually abandoned because of the complexity of processing. To the committee’s knowledge, no waste processing programs are pursuing this technology today.

Modeling has shown that cements can be “designed” to retain radioactive and hazardous constituents (McPhee and Glasser, 1993). In fact, much research has focused on improving the effectiveness of grout in adverse environments associated with the disposal of radioactive waste (Ghattas et al., 1998; Mattus, 1998; Merz and Khalil, 1992). As discussed in these references, a variety of cement-polymer composites have been investigated as a means of making grouts more compatible with the radioactive and chemical constituents in waste.

For example, the addition of blast furnace slag to Saltstone,⁷ which is being used to dispose of low-activity waste at SRS (see Chapter 2), provides a chemical reductant [iron(II)] and a precipitating agent (sulfide) that chemically binds contaminants such as chromium and technetium as insoluble species, thus reducing their tendency to leach from the waste form. Experimentation has shown that leaching of chromium and technetium was effectively reduced to levels that would allow all projected future salt solution compositions to be processed into Saltstone (MMES et al., 1992). These experiments showed that the addition of slag essentially stopped technetium-99 leaching, although it did not reduce nitrate leaching.

3.3.6 Geopolymers and Hydroceramics

Less mature technologies have been investigated for low-activity waste disposal since the mid-1990s. These include geopolymers and hydroceramics, which are described in the following subsections.

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⁷ The Saltstone that is being used to immobilize low-activity waste at Savannah River contains 5 weight percent cement, 25 weight percent fly ash, 25 weight percent blast furnace slag, and 45 weight percent salt solution.

3.3.6.1 Geopolymers

Geopolymers are ceramic-like, inorganic polymers made from aluminosilicates cross-linked with alkali metal ions (Barbosa et al., 2000; Davidovits, 1994; Kriven et al., 2004). A low water content is used (H₂O/M₂O ~10-25 weight percent, where M denotes the alkali metals sodium, potassium, or cesium) so that an amorphous geopolymer forms instead of crystalline zeolites. A nominal composition of 4SiO₂•Al₂O₃•M₂O is often used to represent the geopolymer matrix, although Si:Al ratios can vary from 1 to 3 depending on the intended application. For cement- and concrete-like applications a ratio of 2:1 is nominally used (Sheppard, 2005). Geopolymers appear to be excellent low-temperature binders and are environmentally more acceptable⁸ than cement waste forms.

Geopolymers are typically made by mixing an aluminosilicate such as metakaolinite (Al₂O₃•2SiO₂) or Class F fly ash with a highly caustic hydroxide solution and an alkali silicate solution of the type (Na,K,Cs)₂Si₂O₅. Geopolymers can be formed under ambient conditions (Kriven et al., 2004) but often are autoclave cured between 80°C-120°C to produce an amorphous, cross-linked, three-dimensional structure. Curing of large-scale monoliths in a steam room at 40°C-45°C has been found to be adequate for stabilization of mining wastes (Davidovits, 1995; Hermann et al., 1999; van Jaarsveld et al., 1996).

Geopolymers and geopolymeric cements, including but not limited to fly ash-based geopolymeric concretes, are candidates for environmental applications, including permanent encapsulation of radioactive waste (Hanzlicek and Steinerova-Bopndrakova, 2006; Hanzlicek et al., 2006) and other hazardous waste (Perera et al., 2005), and also as sealants, caps, barriers, and other structures necessary for remediation of contaminated sites. Geopolymers have been used in pilot-scale demonstrations on both mining wastes and uranium mill tailings in Europe (Davidovits, 1995; Hermann et al., 1999; van Jaarsveld et al., 1996) and were also investigated in the mid to late 1990s for the disposal of radioactive wastes (Khalil and Merz, 1994; Zosin et al., 1998).

More recently, geopolymers have been investigated as a waste form for stabilization of Resource Conservation and Recovery Act metals; as binders for granular mineral wastes produced by fluidized bed steam reforming for low-activity waste at Hanford; and for the stabilization of cesium-137 and strontium-90 wastes.

Special geopolymer formulations, marketed under the name DuraLith,

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⁸ Production of the raw materials used to make cement emits CO₂. Production of geopolymer materials emits only water vapor; moreover, geopolymers can be made from fly ash, a byproduct and waste from coal-fired power plants.

have been patented (Gong et al., 2006) by the Vitreous State Laboratory (VSL) in Washington, D.C. Testing of DuraLith formulation by Pacific Northwest National Laboratory (Gong et al., 2006) and VSL (W. Lutze, VSL, written communication) showed great promise for stabilization of technetium.

3.3.6.2 Hydroceramics

Hydroceramics are concrete-type materials that are made by curing a mixture of inorganic waste, calcined clay (metakaolin, Al₂O₃•2SiO₂), vermiculite, Na₂S, and NaOH with water under hydrothermal conditions (60ºC-200ºC) to form a matrix containing crystalline tectosilicates (zeolites) embedded in a sodium aluminosilicate matrix (Bao et al., 2004). The solidification process occurs as a result of hydration reactions. The NaOH solution dissolves the metakaolin much the same as in geopolymers, but the presence of abundant water or hydroxide results in the production of crystalline silicates instead of an amorphous matrix. These silicates have sodalite and cancrinite structures, which can trap constituents and render them insoluble.

Hydroceramic waste forms have several potential applications to DOE-EM waste streams. They have been shown to be effective for immobilizing low-activity sodium-bearing waste at INL. Additionally, Scheetz and Olanrewaju (2001) have developed hydroceramic waste forms for HLW calcines at INL. However, the need for denitration and high-temperature curing will likely preclude the use of hydroceramics for immobilizing low-activity waste streams at SRS and INL.

3.3.7 Ceramicretes

Ceramicretes are phosphate-bonded ceramics, also known as chemically bonded phosphate ceramics. These materials are produced by reacting magnesium oxide, monopotassium phosphate, and the stoichiometrically required amount of water according to the reaction

MgO + KH₂PO₄ + 5H₂O → MgKPO₄•6H₂O

This reaction is exothermic and occurs rapidly at room temperature. The reaction product, ceramicrete (MgKPO₄•6H₂O), is a hard, insoluble phosphate ceramic (Wagh et al., 2003) that contains a considerable amount of bound water. Waste constituents react with the material to form insoluble phosphates or are encapsulated in the matrix.

A patented technology for producing ceramicretes (Wagh and Singh,

1997) has been licensed to treat mixed and low-level wastes and is being used for macroencapsulation and containerization of uranium. Phosphate ceramics are also used for road and highway repairs, and the oil industry is testing these materials for drilling casing and capping. The medical/dental industry is also using several phosphate-ceramic formulations.

DOE has also examined the suitability of this waste form for both micro- and macro-encapsulation of radioactive and hazardous waste streams, e.g., low-activity waste streams typical of Hanford and other sites. Testing of ceramicrete made with INL’s sodium-bearing waste and Hanford Waste Treatment Plant Secondary Waste has been performed at Pacific Northwest National Laboratory.

3.4 DISCUSSION

As the discussion in this chapter illustrates, there are a wide variety of waste form materials that could potentially be used to immobilize radioactive and chemically hazardous wastes. Some waste form materials have been demonstrated in real-world applications on radioactive and hazardous wastes, whereas other have only been researched or demonstrated in the laboratory or at pilot scales.

The committee judges that there are many potential applications of new and improved waste form materials to DOE-EM waste streams. Table 3.7 illustrates the potential compatibility of the waste form materials described in this chapter to some of the DOE-EM waste streams that were described in Chapter 2. Note particularly:

• All eight classes of waste form materials that are described in this chapter are potentially compatible with one or more DOE-EM waste streams.
• Borosilicate glass is already being used by DOE-EM to immobilize HLW, and DOE-EM plans to use glass to immobilize low-activity waste at Hanford (see Chapter 2 and Appendix C). However, other waste form materials, in particular GCMs and crystalline ceramics, are potentially suitable for both of these applications.
• Glass, GCMs, and crystalline ceramic materials have the widest range of potential compatibilities with DOE-EM waste streams—in particular for HLW from tanks and high-sodium wastes, but also for cesium and strontium capsules (should DOE-EM decide not to dispose of those capsules directly) and excess plutonium.
• One or more of the encapsulant waste form materials (i.e., cements, geopolymer, ceramicrete, and hydroceramic waste forms) are potentially compatible with high-sodium wastes, cesium and strontium capsules, and wastes containing iodine-129 and technetium-99.

TABLE 3.7 Waste Form Compatibility with Selected DOE-EM Waste Streams

NOTES:

C = current application; P = possible application

^a With blending, dissolving, low waste loading.

^b Pu solubility low.

^c With Bi-P or high SO₄.

^d High waste loading.

^e With high PO4, high waste or high Cr.

Of course, compatibility is just one of several considerations in selecting a waste form material to immobilize a specific waste stream. Other considerations include the ease of processing, cost, and risk. Processes for waste form production are described in the next chapter. The committee’s finding on the fourth charge of its task statement (Box 2.1. in Chapter 2) is given in Chapter 1.

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