Alternative High-Level Waste Treatments at the Idaho National Engineering and Environmental Laboratory (1999)

Vitrification has a long track record as a process for immobilizing nuclear waste and is being used at national facilities such as the Savannah River Site and West Valley Demonstration Project. At the Idaho National Engineering and Environmental Laboratory (INEEL), research on waste vitrification has been conducted since the 1960s. The method consists of mixing the high-level waste (HLW) with glass powders (''frit"), melting the mixture at high temperature (e.g., 1150 °C)¹ in a joule-heated melter lined with ceramic brick,² and pouring the melt into waste containers. After cooling, the resulting nuclear waste glass products are generally homogeneous, noncrystalline materials with high chemical durability. They are essentially free from an undissolved crystalline phase but can contain some precipitated crystalline phases and separated amorphous phases as minor phases to the extent that their overall chemical durability is not affected adversely.

This chapter begins by evaluating the vitrification plans for the INEEL HLW and associated developmental studies that were contained in presentations to the committee and in various technical reports cited below. The nonseparation option avoids chemical separation steps by vitrifying the solid calcine directly. The separation option vitrifies the high-activity waste streams resulting from calcine dissolution (including undissolved solids; see Chapter 2) followed by chemical separations processing (see Chapter 3). These two options, and the status of developmental testing conducted to support them, are discussed below in greater detail. Both of these options use continuous, joule-heated melters and borosilicate-based glass compositions.

The chapter ends with a discussion of potential problems for these vitrification approaches, and mentions possible solutions. A fuller treatment of technological alternatives is deferred to Chapter 7, following discussion in Chapters 6-7 of other immobilization approaches. That discussion does not summarize all emerging and viable vitrification techniques such as plasma furnaces or induction-heated cold crucible melters that can be considered for highly refractory materials, but focuses on some developments that in the committee's view were worthy of further consideration for application on INEEL HLW.

In the nonseparation option, HLW calcine is directly mixed with glass frit and convened to glass. At present, there are approximately 4,000 m³ of HLW calcine at INEEL, which consists of alumina-and zirconia-based calcine and zirconia-sodium blend calcine (Knecht et al., 1996). In addition, if ail existing and projected future liquid wastes are solidified, approximately 2,000 m³ of additional calcine will be produced (Palmer, 1996) primarily from sodium-bearing waste (SBW). In this option, it is estimated that a total of 14,000 "Savannah River-size" canisters (approximately 0.7 m³ each) will be filled with glass with a waste loading of approximately 20 to 30 percent by weight.

The base glass composition developed for use with Savannah River and West Valley HLW is a borosilicate type, with the relative abundance of frit constituents tailored to the waste composition to ensure production of a vitrified form with adequate waste loading. However, the calcine compositions at INEEL are quite different from the waste compositions found at these other facilities, in that the INEEL HLW calcines contain significantly higher concentrations of chemicals that have an important effect on glass formation and properties. For example, the alumina-based calcine contains 82 to 95 percent by weight Al₂O₃, the zirconia-based calcine contains 21 to 27 percent by weight ZrO₂ and 50 to 56 percent by weight CaF₂, and the SBW contains high concentrations of sodium nitrate. These waste compositions affect the glass formulation required to achieve a high waste loading. Accordingly, new nuclear waste glasses have to be designed to accommodate these unique compositions. These glass formulation efforts at INEEL are summarized next.

In the 1970s and 1980s, research was performed at INEEL to develop waste glass compositions for zirconia-based calcine. A borosilicate glass frit "127" with the composition SiO₂ 70.3, Na₂O 12.8, Li₂O 6.2, B₂O₃ 8.5, CuO 2.1 percent by weight was developed. With 25 to 35 percent by weight calcine and 65-75 percent by weight frit, the mixture was melted at 1100 °C for 3 hours (Staples et al., 1983). Glasses were prepared both on laboratory and pilot-plant scale using simulated nuclear waste calcine. Radioactive waste calcine was also used in laboratory-scale melting. The zirconia content of the resulting glasses was approximately 9 percent by weight (Staples et al., 1983). The durability of these glasses, as tested in the materials characterization center (MCC) tests MCC-1 and MCC-2, was comparable to that of the Savannah River glasses. There was no significant difference in the glass characteristics between samples prepared in laboratory scale and pilot-plant scale as well as between simulated and radioactive calcine.

The glass developed for the zirconia-based calcine cannot be used for the alumina-based calcine nor for a mixture of zirconia-based calcine and alumina-based calcine with more than 15 percent by weight alumina calcine (Brotzman, 1978). Viscosity measurement and Soxhlet leach test of phospho-borosilicate glasses with 23 mole percent (approximately 31

percent by weight) alumina-based waste loading were tested to develop a satisfactory nuclear waste composition that can be melted at 1100°C (Brotzman, 1978). The base composition used was SiO₂ 3.2, B₂O₃ 31.3, P₂O₅ 18.9, Li₂O 4.0, Na₂O 8.3, CuO 3.2, and Al₂O₃ calcine 31.1 percent by weight. A subsequent report (Cole and Colton, 1982) indicated that for alumina-based calcine, the borosilicate glass "532" with the composition SiO₂ 46.0, B₂O₃ 10.0, Na₂O 7.0, Li₂O 7.0, CuO 2.0, TiO₂ 5.0, and alumina-based calcine 23.0 percent by weight, gave the best durability among several tested glass compositions.

Glasses were developed (Vinjamuri, 1995) for directly vitrifying the SBW. The flit consists of borosilicate glasses with high (>80 percent by weight) silica content. A glass was prepared with a waste loading of more than 20 percent by weight, and its durability was found to be satisfactory. Specifically, the leach rate normalized to glass composition was low.

Glasses were developed to vitrify a waste consisting of a mixture of alumina calcine and SBW. In particular, blends of pilot-plant alumina calcine and SBW can be immobilized using the same high-silica borosilicate glass compositions that were developed for the SBW only (Vinjamuri, 1995). For example, a blend of the simulated waste consisting of 57 percent by weight alumina calcine and 43 percent by weight SBW was mixed with a flit consisting of 88 percent by weight SiO₂ and 12 percent by weight B₂O₃ with the waste loading ranging from 20 to 35 percent by weight. The mixtures were melted at 1200 °C or 1600 °C³ for 2 to 5 hours and their durability measured by the MCC-1 leach test. Glass "532" developed for the alumina-bearing calcine is reported (Cole and Colton, 1982) to be satisfactory for the blend consisting of alumina-based calcine and SBW in the ratio of 2.5:1.

In addition to the development of adequate glass compositions to accommodate the HLW calcine, other developmental challenges are associated with the design, construction, testing, and operation of a full-scale vitrification plant to operate on radioactive waste feed. INEEL has made detailed plans to build a facility to vitrify calcine waste (Lopez and Kimmitt, 1998). The operation includes blending of the calcine, sampling of the calcine to determine its composition, mixing the calcine with an appropriate glass flit selected from six different compositions, delivering the mixture to the melter, melting and pouring the glass into a metal canister, processing of off-gases, and finally transporting the canister to the interim storage facility. According to this study, the planned facility has a capacity to produce, on average, 3.9 "Savannah River size" canisters per day filled with glass.

In the separation option, calcined waste is dissolved and then separated into high-activity waste (HAW) and low-activity waste (LAW), and the HAW portion is vitrified (Murphy et al., 1995). The incentive for the separation option is the reduction in the vitrified waste volume, which is estimated to be less than one-tenth of that without separation, with the number of canisters expected to be 1,100 instead of 14,000 (Murphy et al., 1995). As a result of the separation process, the HAW will contain a high concentration of zirconia or of potassium and phosphate. For example, the ZrO₂ content in HAW produced from zirconia-based calcine is estimated to be 95 percent by weight, and the K₃PO₄ content in HAW produced from alumina-based calcine and sodium-bearing calcine is expected to be 75 percent by weight and 85 percent by weight, respectively (Staples et al., 1998). These unique waste compositions require special consideration in developing host glasses that can contain reasonable amounts (approximately 25 percent by weight) of these compositionally unusual wastes.

As with the nonseparations option, development work has been initiated on glass formulations that would be suitable for the compositions of waste to be vitrified. Because the above-mentioned HAW compositions resulting from the separation process are unconventional, a group effort was instituted to develop appropriate nuclear waste glass compositions (Staples, et al., 1998). The group consisted of personnel from INEEL, Pacific Northwest National Laboratories (PNNL), and the Savannah River Technology Center (SRTC).

This group developed and tested glasses designed to immobilize the compositions of HAW that are estimated to be generated from processing the various types (e.g., zirconia and alumina) of INEEL HLW calcines and SBW. Various borosilicate glass compositions were melted at 1,150 °C for 4 hours and their properties (e.g., viscosity, liquidus temperature, and chemical durability) were evaluated with the objective of producing a nuclear waste glass that can accommodate at least 19 percent by weight of the "all-blend" HAW waste composition. The chemical durability of some of these glass compositions, as determined by the product consistency test (PCT), was found comparable W the Environmental Assessment (EA) glass. Phase separation was observed in the nuclear waste glasses when the phosphate concentration exceeded 5 to 7 percent by weight. In the borosilicate glasses studied, the waste loading was limited by the zirconia and/or phosphate content of the final waste glass composition. Upper limits of approximately 15 percent by weight zirconia and/or 3 to 5 percent by weight phosphate were derived for the final waste glass composition (Staples et al., 1998). This work is still in progress.

At the present time, there is no detailed plan to build a vitrification facility to process the HAW, but a scaled-down version of the vitrification facility for the nonseparation option presumably could be used.

As noted in Chapter 2, the calcine compositions in the bin sets vary widely, containing large amounts of components like ZrO₂, Al₂O₃, and CaF₂ that are only sparingly soluble in common borosilicate glasses, or components like P₂O₅ that cause phase separation in these same glasses. These calcine compositional variations and chemical constituents pose potentially serious problems for achieving high waste loading in borosilicate glass, as discussed below.

The retrieval and blending during retrieval (Chapter 2) of calcines of different compositions raises the issue of how to achieve a sufficiently uniform composition of a waste stream for a vitrification process. The waste stream in view here is either undissolved solid calcine, or calcine dissolved in acid solution with some constituents removed. One solution is a design that blends large enough quantities of different types of calcine (after separations steps, if any), by mixing them in a large "feed tank," and to design a frit composition that is compatible with the feed tank calcine composition. A quasi-batch mode would use successive vitrification campaigns to work off successive feed tank volumes; alternatively, if the feed tank were large enough to provide for on-line compositional monitoring of the feed and subsequent adjustment of frit composition, a continuous mode of operation might be possible.

In this approach, an appropriate glass frit (or individual oxide components) can be chosen from among several candidate compositions developed for this purpose. The compositional adjustments to the frit would be minor if the compositional variations of the blended calcine feed were likewise minor. However, the compositional changes in frit could be large if the compositions of the successive blends of calcine are considerably different. Since there are several bins to choose from in recovering the calcine, the compositional variations from blend to blend can be minimized by proper choice of the bins to be emptied. These variations would be prudent to plan for, in the absence of sufficiently large-scale blending to produce a uniform calcine mixture, because of the differences in calcine types stored within each bin.

Since alumina and zirconia are only slightly soluble in borosilicate glass at the planned melting temperature of 1150 °C, calcines rich in these oxides would have a low solubility in glass and a relatively lower (probably 15 to 20 percent by weight) waste loading than the 27 percent by weight achieved at Savannah River. If separations alter the content of species such as zirconia, the waste loading would be correspondingly affected.

In the mid-1960s, a phosphate glass was studied for possible use as a nuclear waste glass. Since 1970, however, only borosilicate glasses have been investigated on a large scale in the United States, with two exceptions: phospho-borosilicate glass compositions for a separation option (Staples et al., 1998) and lead-iron phosphate glass development (Sales and Boatner, 1988). In view of the unique composition of the various wastes at INEEL, especially those of high P₂O₅ concentration, new phosphate glass compositions deserve attention (Day et

al., 1998), because such phosphate glasses could contain larger mounts of these unique wastes and have equally good chemical durability. While borosilicate-based nuclear waste glasses appear to exhibit phase separation that leads to chemical durability deterioration when the P₂O₅ content exceeds 5 to 7 percent by weight (Staples et al., 1998), depending on the final waste form composition, phosphate-based glass should be able to avoid this problem. Therefore, further examination of phosphate-based compositions would be useful to probe issues of chemical durability and corrosion.

Zirconia has a limited solubility in oxide glasses and slow dissolution kinetics, especially at the melting temperature of 1150 °C proposed for nuclear waste vitrification. Therefore, process options other than vitrification may be more expedient ways to handle zirconia-rich waste. In the production of a homogeneous vitrified waste form, the zirconia would be dissolved completely. However, zirconia may be more useful in a crystalline form serving as a host for some radioactive elements. Indeed, there is a growing literature (e.g., Heimann and Vandergraaf, 1988; Oversby et al., 1997) on the use of zirconia as an inert fuel matrix and durable waste form. A glass/ceramic, made by melting or sintering, may be a better waste form for zirconia-rich materials than a vitrified waste form, since a higher waste loading is expected (resulting in fewer canisters), and it should have an acceptable chemical durability.

The calcium fluoride component of the feed to a vitrification process is potentially problematic. The concern regarding fluorides such as calcium fluoride stems from their tendency to vaporize at high temperatures (which results in the possible formation of corrosive off-gas constituents containing fluorine), their limited solubility ⁴ in glasses such as borosilicates, and their tendency to promote phase separation (which has the potential to reduce chemical durability or increase the rate of refractory or electrode corrosion) in borosilicate glasses. These potential problems can be avoided by efforts such as (1) removing the fluorides such as calcium fluoride from the waste prior to vitrification,⁵ (2) designing the glass composition to accept a small amount of fluoride content,⁶ or (3) using glasses other than borosilicates that are more compatible with fluorides. Fluorides are typically more soluble in phosphate glasses than in borosilicate glasses; as an example, fluorophosphate glasses are well known in the optical glass field and have been manufactured commercially for several decades.

If Joule-heated melting is chosen for the INEEL HLW calcine, prototype testing will be needed. Important issues that should be addressed would include corrosion of the refractories and electrode materials, atmosphere and off-gas controls,⁷ and facilities for draining the melt into canisters and/or for removing insoluble materials (e.g., noble mews) from the furnace floor. In the absence of waste composition data, these required tests cannot be specified more definitively. A minimum requirement, however, should be that corrosion tests with candidate refractories and electrode materials be performed using simulated, if not actual, calcine waste, to exhibit behaviors important to the design of a full-scale process.⁸