6
Cementation

Cement-based mixtures have a long history of use for immobilization of low-level radioactive wastes (Moore et al., 1975; Moore, 1981). The conventional immobilization process incorporates waste in a mixture of portland cement, clays, bases, and water. Upon drying or thermal treatment, mineral phases such as feldspathoids, zeolites, and other hydroceramics are formed that contain the waste iota. The waste form is durable and resistant to leaching by water.

The procedures and equipment to produce cementitious waste forms are modifications of practices used to produce ordinary concrete. Mixers, forms (for shaping), and drying and curing equipment would be used in remote or shielded operations.

Concretes, such as those "formed under elevated temperature and pressure" (FUETAP) that were developed at the Oak Ridge National Laboratory (McDaniel and Delzer, 1988), also have been proposed for the immobilization of high-level wastes (HLW). FUETAP utilizes the thermal output of the waste to accelerate the curing process. Development work at the Idaho National Engineering and Environmental Laboratory (INEEL) has built on the FUETAP experience in the study of cementitious waste forms for the immobilization of INEEL HLW (Russell et al., 1998; Dafoe and Losinski, 1998; Lee and Taylor, 1998). These forms have been considered for use on the following wastes:

  • the high activity waste (HAW) generated after several separations steps have been performed on the dissolved calcine or sodium-bearing waste (SBW),

  • the HAW generated after only one transuranic (TRU) separations step has been performed on the waste stream, and

  • the HLW with no separations steps at all; that is, the solid calcine and liquid SBW.

Each of these separations-based and nonseparations-based options have different processing parameters, give different end products, and require different ultimate disposal options. For example, some of these options generate a low-activity waste stream, which is also immobilized in a cementitious waste form.

A technical assessment of the various cementitious processes and end products with respect to the given statement of task was performed and is described in the paragraphs that follow. The cementation processes, described first, are relatively straightforward operations as compared to other immobilization techniques in Chapters 5 and 7. The quantity of final waste



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Alternative High-Level Waste Treatments at the Idaho National Engineering and Environmental Laboratory 6 Cementation Cement-based mixtures have a long history of use for immobilization of low-level radioactive wastes (Moore et al., 1975; Moore, 1981). The conventional immobilization process incorporates waste in a mixture of portland cement, clays, bases, and water. Upon drying or thermal treatment, mineral phases such as feldspathoids, zeolites, and other hydroceramics are formed that contain the waste iota. The waste form is durable and resistant to leaching by water. The procedures and equipment to produce cementitious waste forms are modifications of practices used to produce ordinary concrete. Mixers, forms (for shaping), and drying and curing equipment would be used in remote or shielded operations. Concretes, such as those "formed under elevated temperature and pressure" (FUETAP) that were developed at the Oak Ridge National Laboratory (McDaniel and Delzer, 1988), also have been proposed for the immobilization of high-level wastes (HLW). FUETAP utilizes the thermal output of the waste to accelerate the curing process. Development work at the Idaho National Engineering and Environmental Laboratory (INEEL) has built on the FUETAP experience in the study of cementitious waste forms for the immobilization of INEEL HLW (Russell et al., 1998; Dafoe and Losinski, 1998; Lee and Taylor, 1998). These forms have been considered for use on the following wastes: the high activity waste (HAW) generated after several separations steps have been performed on the dissolved calcine or sodium-bearing waste (SBW), the HAW generated after only one transuranic (TRU) separations step has been performed on the waste stream, and the HLW with no separations steps at all; that is, the solid calcine and liquid SBW. Each of these separations-based and nonseparations-based options have different processing parameters, give different end products, and require different ultimate disposal options. For example, some of these options generate a low-activity waste stream, which is also immobilized in a cementitious waste form. A technical assessment of the various cementitious processes and end products with respect to the given statement of task was performed and is described in the paragraphs that follow. The cementation processes, described first, are relatively straightforward operations as compared to other immobilization techniques in Chapters 5 and 7. The quantity of final waste

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Alternative High-Level Waste Treatments at the Idaho National Engineering and Environmental Laboratory that is acceptable to produce and the qualification of the cementitious waste form for suitable disposal are the major issues that would need to be resolved to make cementation a fully viable option. The waste form qualification issues are treated at the end of this chapter, and the issue of the quantity of high-level waste is taken up in Chapters 9 and 10. PROCESS DESCRIPTIONS Brief process descriptions of the direct (nonseparations) and separations cases for the waste materials presently at INEEL are given below. Two direct and three separations-based cementation processes have been considered. Direct Cementation: Mixed HLW Hydroceramic with Feed of Clay, Slag, Soda, and Water In this option, the existing SBW liquids would be calcined and added to the other HLW calcine presently in the Calcined Solids Storage Facility (CSSF) at INEEL. This mixed calcine would then be mixed with clay, blast furnace slag, caustic soda, and water such that analogs of naturally occurring feldspathoids/zeolites are generated. This cement/waste mixture would be extruded into stainless steel canisters where a curing process would produce a structurally sound and geologically stable hydroceramic waste product. The canisters used would be the same as those used at the Defense Waste Processing Facility at the Savannah River Site (Dafoe and Losinski, 1998). This process does not minimize the volume of waste to be disposed of and relies on furore acceptance of the mixed HLW form (i.e., containing both hazardous chemicals and radionuclides) at a currently unapproved site such as the Greater Confinement Disposal Facility (GCDF)1 at the Nevada test site (e.g., NTS) (see Chapter 9). That is, this option does not involve the site of the geologic repository for spent nuclear fuel and co-disposed defense HLW. The calcining and cementation would be accomplished within a 20-year operating cycle (Dafoe and Losinski, 1998). The process outlines have been given (Dafoe and Losinski, 1998) and constitute straightforward processes and process equipment. This process could be implemented simply, with little development, and it has a low risk of failure (Dafoe and Losinski, 1998) and minimum personnel exposure. As a result, costs should be relatively low compared to processes requiring redissolution of existing calcine and various separations processes. It should be possible, due to process simplicity, to shorten the stated 20-year operating cycle quoted in Dafoe and Losinski (1998). Direct Cementation: Mixed HLW Hydroceramic with Feed of Sucrose, Clay, NaOH, and Water The second case of direct cementation is described by Lee and Taylor (1988) and involves different feed materials than the first case described above. In this case, SBW liquids would be slurried with calcine from the CSSF and sucrose and recalcined. This would require the existing calciner to be repermitted. This new calcine would be classified as mixed HLW. The new calcine would then be mixed with calcined kaolinite clay, sodium hydroxide, and 1   The GCDF is a series of boreholes in arid alluvium of the Nevada Test Site that have been used in the past to store Department of Energy (DOE) classified TRU waste (see Chapter 9).

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Alternative High-Level Waste Treatments at the Idaho National Engineering and Environmental Laboratory water. The cement mixture would be placed into HLW canisters, steam Cured via autoclave, dewatered, degassed, and sealed. These would then be placed in interim storage for transfer to a packaging facility and transport to the GCDF (Lee and Taylor, 1988). The cementitious waste form produced is a hydroceramic designed to be stable in the NTS alluvium. This process does not minimize waste volumes but does use simple and inexpensive feed materials. The process and process equipment are simple, easy to operate and commonly available. The process does rely on future acceptance of a mixed HLW waste form at an as yet unapproved site (e.g., NTS). The calcine and SBW can be processed using this approach in about 5 years (Lee and Taylor, 1988). The process outlines for this option have been given (Lee and Taylor, 1988) and are again straightforward. Because of process simplicity and simple equipment, the risk of failure is low and personnel exposure should be low. As a result, costs should be low relative to those processes requiring calcine redissolution and various separations. This option does depend on the existing or similar calciner being usable soon and probably entails the lowest risk and lowest development effort of any of the various cementitious processes. In both of the two direct cementation processes described above, the resolution of the treatment and disposal of all Resource Conservation and Recovery Act (RCRA) hazardous materials must be pursued energetically because it may be difficult to dispose of mixed HLW. In addition to the RCRA problem, there would be approximately 13,000 cubic meters of cemented waste formed with the first process (Dafoe and Losinski, 1998) and approximately 12,000 cubic meters with the second process (Lee and Taylor, 1988). Although neither of these volumes should be a problem if the GCDF is made operable, the committee believes any planned use of a GCDF must recognize that the use of such a disposal site for INEEL HLW is in a very early stage of conceptual development and outside the current regulatory approach (see Chapter 9). Separations-Based Cementation Three processing options, briefly discussed below, involve separating specific components from the HLW and SBW to decrease the amount of HLW requiring disposal at a HLW repository. The bulk of the waste volume remaining after these separations are performed would be grouted into a low-level waste (LLW) form. TRU Separations with Class C Grout In this option, TRU components are separated for disposal at the Waste Isolation Pilot Plant (WIPP) and the resulting LLW waste would be grouted for disposal as Class C waste (Option 2.2.2 of Russell et al., 1998). Further details are provided in Landman and Barnes (1998). The TRU separation processes have only undergone small-scale demonstrations at INEEL. The major advantages of this option are (1) the use of only TRU separations (i.e., Cs and Sr separations are avoided), (2) a single off-gas treatment facility, and (3) the WIPP repository, which is potentially available in the future to accept mixed TRU waste. A regulatory reclassification ruling would be required because this option proposes to produce no HLW outputs.

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Alternative High-Level Waste Treatments at the Idaho National Engineering and Environmental Laboratory Haw and TRU Separations with Class a Grout This option involves separation of two fractions—HAW and TRU—from the LLW stream (Option 2.2.3 of Russell et al., 1998). The HAW fraction, containing both cesium and strontium, would be processed into a vitrified HLW product. As in the previous option, the separated TRU fraction would be dried and packaged for disposal at WIPP. After separations, the LLW would be concentrated by evaporation, denitrated, and grouted as a Class A waste. This option requires significant dissolution and separations steps that have not been demonstrated in pilot-plant level operations. A regulatory reclassification ruling would be required, as above, to approve of a TRU waste stream derived from a HLW input. Separate TRU, Cs, and Sr Separations with Class a Grout The third separations option (Option 2.2.1 of Russell et al., 1998) separates three fractions—TRU, Cs, and St—from the LLW stream. These three separately generated fractions are then combined and vitrified into a HLW product. The remaining LLW would be grouted after denitration for disposal as a Class A waste. Again, further details are provided in Landman and Barnes (1998). The processing disadvantages and regulatory issues are similar to those of the previous two separation processes. Because Class A grout has fewer disposal restrictions than Class C grout, the Class A LLW grout of this and the previous option are relatively easier to dispose of, although the RCRA constituents of the grout likely represent a more important restriction that all options need to address. In these last two options, an interim-storage facility would have to be built at the INEEL to store the vitrified HLW until a repository became available for its disposal. Disposal of LLW Grout The three separations-based operations all produce a large volume (approximately 22,000 to 27,000 m3) of LLW grout to be disposed. For this LLW grout, three specific disposal concepts have been considered (Russell et al., 1998). In one, grouted LLW would be pumped into the empty storage tanks and the empty bin sets. In the second, grouted LLW would be packaged in thick-walled, cubic-meter-size concrete containers suitable for disposal in a near-surface facility. In the third, grouted LLW would be packaged and sent to the Hartford site in the State of Washington for disposal. While the use of the liquid storage tanks and calcine bins for storage of a large volume of cementitious LLW is technically and economically attractive, all of these concepts depend on delisting some of the RCRA wastes (see Chapter 9). Because of the relative simplicity of cementation equipment and processes, these options for immobilizing INEEL HLW are attractive. No high temperatures are required, and in some options, no chemical separations are required. These conditions lessen hazards to workers and the public. However, because of the many possible mineral compounds formed, depending on the starting materials and the waste constituents, significant further testing is required. This issue is taken up in the next section.

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Alternative High-Level Waste Treatments at the Idaho National Engineering and Environmental Laboratory CEMENTITIOUS WASTE FORMS The cementitious waste form proposed for use on INEEL HLW is described in a recent paper by Siemer et al. (1998). The waste form is not the conventional portland cement-based waste form, but rather is of a special composition based on the alkaline activation of pozzolanic aluminosilicates. In this waste form, aluminosilicates, such as the zeolites (Siemer et al., 1998), are important radionuclide-bearing phases. The approach builds on previous experience with FUETAP and uses a ''mild" hydrothermal processing (identified as a "clay reaction process"). The waste form was prepared by having kaolintic clay react with a range of simulants for the SBW. Samples were cured at approximately 200 °C for several hours. Samples were characterized by x-ray diffraction (XRD) and scanning electron microscopy (SEM), and three types of leach tests [the standard leach tests included the ANSI-16.1 standard, the Product Consistency Test (PCT), and the Toxic Characteristic Leaching Procedure (TCLP)] were performed on the product. Principal phases resulting from the synthesis included unreacted quartz, a hydroxysodalite, and zeolite-A (Na12Al12Si12O48 27H2O). The results of the leach tests are interpreted to show that the cementitious waste form provides equivalent or better performance than typical waste form glasses (despite a much higher surface area for the cementitious waste form). No specific data were presented for the waste form products that would result from the two different processing options of direct cementation and separations-based cementation. It is clear from the literature on cement that a wide variety of phases may form depending on initial compositions and processing parameters. Despite the considerable literature on cementitious waste forms and the fact that cementation is a candidate technology for INEEL waste, it does not appear that much recent work has been focused on the development and characterization of the cementitious waste form. Indeed, the justification for the use of the cementitious waste form, aside from the ease of processing, is the following: (1) it is comparable to glass in its performance and/or (2) the waste form is of little consequence in the overall performance assessment of a geologic repository. Setting these issues aside, however, there is another major issue, with a number of sub-issues, that should be addressed if the cementitious waste form is to be thoughtfully compared to the alternatives (e.g., vitrification, glass ceramic, sintered glass, or calcine). The major issue is that there are inadequate data to support the contention that cementitious waste forms can be made more quickly, more cheaply, more simply, and more safely than other (e.g., vitrified) waste forms. That contention may be tree, but it should be established by thoughtfully designed experimental programs and an analysis of comparable data (a full-scale demonstration of the process is not required). Such experimental programs and comparable data analyses would include the following: It would be necessary to develop laboratory scale synthesis procedures that reflect both the compositions of materials that will actually be used and the process technologies that will be developed and compared. There is a surprising lack of INEEL laboratory data on the potential synthesis processes and the final products. Systematic studies should be completed to investigate the effects of variations in the composition of the waste stream, variations in waste loading, and the consequent effects on product consistency. A complete and precise characterization of the cementitious waste forms would have to be done. The present use at INEEL of XRD and SEM, as illustrated in Siemer et al. (1998), is cursory at best. Phase identification is tentative, and neither XRD nor SEM will detect many minor but important radionuclide-bearing phases. The characterization must be completed as a function of the initial starting material compositions and over the range of synthesis conditions (e.g., variations in temperature and curing time). It is important that standard, well-described procedures lead to a consistent product. It is also important to establish

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Alternative High-Level Waste Treatments at the Idaho National Engineering and Environmental Laboratory the effect of variations in SBW and calcine compositions on the final product. The characterization must include not only the identification of phases, but also the descriptions of the micro-structure and surface area and the partitioning of radionuclides into particular phases. This is exactly the information that will be required for any substantive assessment of waste form performance as a function of leaching under different geochemical conditions or in a radiation field. A more fundamental understanding of leaching mechanisms is needed. The three leach tests (ANSI-16.1, PCT, and TCLP) used in previous studies provide only a qualitative basis for comparison of waste form performance. Longer term tests with more complete analysis of the solution chemistry, followed by solid-state characterization of the waste form after leaching, are required. The release, transport, and precipitation of radionuclides may occur over very short distances and result in important changes to the material form that could affect long-term performance. It is also important to establish the stability of zeolitic phases in the cementitious waste forms and in the proposed repository environment. In addition to leach tests to measure durability, it would be necessary to determine the effect of the high level of dissolved salts that accompany radioactive liquid wastes on the durability of the waste form. Recent work (Guerrero et al., 1998) has demonstrated that there can be important effects related to the formation and evolution of reaction products (e.g., brushite, a hydrated calcium phosphate). The formation and dissolution of these reaction product phases can affect the porosity and surface area of the cementitious waste form and the subsequent release of radionuclides. Additional study of the effect of radiolysis of structural water in the hydrated phases in the cementitious waste form is necessary. Radiolysis can lead to the formation of gases and subsequently affect the porosity and mechanical strength of the waste form. This is a phenomenon that has received only limited attention; however, the solid-state radiolysis of compounds that contain structural water may seriously affect the long-term durability of a cementitious waste form.2 Also required are additional data on the effect of mercury, an important component of some of the wastes. There must be an investigation of the fate of Hg during processing and in the final waste form in order to predict the amount, chemical form, and long-term behavior of any Hg constituents. SUMMARY As can be seen from the foregoing, the proposal to use a cementitious waste form, particularly one in which zeolites are used to advantage to retain critical radionuclides (e.g., Cs) may have merit. A principal advantage is that the SBW and/or the calcined waste might be directly incorporated into the cementitious waste form with minimal or no treatment. However, an informed decision on the use of a cementitious waste form must be based on a thorough analysis of experimental data gathered on waste forms typical of the selected processing technology. At present, these data do not appear to be available. In addition, the analysis and presentation of the data that are available are not in a form that allows for an appropriate analysis. 2   In the event that gas generation (through radiolysis or other processes) were to be a problem, the waste form containers would have to be engineered to prevent overpressurization, such as by using recombiner catalysts, which was done in shipments of Three Mile Island core debris to INEEL.