Click for next page ( 120


The National Academies | 500 Fifth St. N.W. | Washington, D.C. 20001
Copyright © National Academy of Sciences. All rights reserved.
Terms of Use and Privacy Statement



Below are the first 10 and last 10 pages of uncorrected machine-read text (when available) of this chapter, followed by the top 30 algorithmically extracted key phrases from the chapter as a whole.
Intended to provide our own search engines and external engines with highly rich, chapter-representative searchable text on the opening pages of each chapter. Because it is UNCORRECTED material, please consider the following text as a useful but insufficient proxy for the authoritative book pages.

Do not use for reproduction, copying, pasting, or reading; exclusively for search engines.

OCR for page 119
5 Waste Form Testing T he third charge of the statement of task for this study (see Box 2.1 in Chapter 2) calls for the identification and description of “state- of-the-art tests and models of waste forms used to predict their performance for time periods appropriate to their disposal system.” This chapter describes waste form testing and the use of test results to inform the development of models for evaluating the long-term (103-106 year) performance of waste forms and their associated disposal systems.1 The application of such models to waste forms and disposal environments is discussed in Chapter 7. In the context of this report, a test is a laboratory procedure for mea- suring short-term (days to months) release rates of radioactive and chemi- cal constituents from a waste form material and the formation of reaction products. It typically involves the leaching of a monolithic or crushed speci- men of a waste form material under carefully controlled conditions. Release rates reflect the durability of a waste form material, that is, its resistance to physical and chemical alteration. A large number of standard test protocols have been established by the American Society for Testing and Materials (ASTM) and other organiza- tions; some of the principal tests that are used to investigate waste form materials for disposal applications are described in this chapter. 1 Box 5.1 provides definitions for a number of specialized terms that are used in this chapter. 119

OCR for page 119
120 WASTE FORMS TECHNOLOGY AND PERFORMANCE BOX 5.1 Key Terms and Concepts Experiment: The application of tests to a waste form material to gain a better understanding of its degradation behavior and the release of radioactive constituents. Dissolution: A process (or processes) by which mass transport from a solid waste form to a liquid takes place (see ASTM C1308, ASTM C1220). Dissolution is the result of mechanistic reactions in which chemical bonds are broken and constituents are released from a material and become solvated in a test solution (see ASTM C1662). Dissolution rate: The rate of mass removal per unit time normalized to surface area of the material. Durability: The resistance of a waste form material to chemical and physical altera- tion and the associated release of contained radioactive and hazardous constituents. Leaching: The loss of radioactive or chemical constituents from a waste form by diffusion or dissolution. Performance: The ability of a waste form (waste form performance) or a disposal system containing the waste form (disposal system performance) to seques- ter radioactive and chemical constituents. Release mechanisms: The process that controls the rate of mass transport out of a specimen during dissolution (see ASTM C1308). Solubility: The thermodynamically limited saturation state or equilibrium concen- tration limit of species in solution. Standard test protocols: A standardized procedure for testing a specific type of material to generate a clearly defined test response. In principle, any test can be applied to any material to generate a response. However, a response will be meaningful only when the test and material are matched appropriately (see Section 5.1). Waste form qualification: Demonstration that a waste form material will have ac- ceptable performance in a specific disposal facility and can be fabricated with acceptable performance control. Waste form test protocols: Standard tests developed by organizations such as the American Nuclear Society, American Society of Testing and Materials, International Atomic Energy Agency, and the International Organization for Standardization (see Section 5.3). 5.1 PURPOSE OF WASTE FORM TESTING Laboratory testing of waste form materials is undertaken for several purposes, including to: • Conduct experiments to elucidate the release mechanisms of radio- active and chemical constituents from a waste form material.

OCR for page 119
121 WASTE FORM TESTING • Control waste form production. • Ensure that production results in an acceptable waste form material. • Provide the information needed to model the performance of waste forms in disposal systems. These applications are described briefly in the following paragraphs. Experiments involve the application of tests to waste form materials to gain a better understanding of their degradation behavior and release of constituents. These experiments may or may not be developed as standard protocols, but rather they are designed to address and challenge specific hypotheses about the behavior of waste form materials. Such experiments can also involve modeling and other types of measurements, for example, compositional analyses of alteration products that are formed on the sur- faces of waste form materials during release. Radioactive and chemical constituents can be released from a waste form material by one or more of the following three mechanisms: • Reaction affinity-controlled release: Release is controlled by the dif- ference in Gibbs free energy between the thermodynamically stable state and the metastable reactants. • Solubility-controlled release: Release is bounded by the use of the maximum saturation of a constituent species from the waste form in the given leachant (solution) environment. • Diffusion-controlled release: Release is controlled by the diffusion of a constituent in the waste form material, including diffusion through an encapsulant and/or through surface layers containing reaction products, if present. In some cases a change in oxidation state of the constituent may occur prior to its release. Laboratory experiments on natural analogues of waste form materials (e.g., basalt glass as a natural analogue for borosilicate glass) allow one to gain insights into the similarities and differences in release mechanisms. Short-term studies of natural analogues can also be extended to investigate other material properties, for example, for comparing radiation damage in actinide-doped materials with damage in uranium- and thorium-bearing minerals (Weber et al., 1994). Testing for production control is used to determine how the production of a waste form material affects (or controls) its performance, and also to identify the ranges of processing variables that produce acceptable waste forms. The primary role of production control testing is to verify that the properties of a specific waste form product are consistent with the waste

OCR for page 119
122 WASTE FORMS TECHNOLOGY AND PERFORMANCE form material deemed to be acceptable for disposal, either by direct mea- surement or through process control. Waste form acceptance testing is intended to show that the waste form produced within production control limits will have acceptable perfor- mance in a disposal facility. The performance of a waste form in a disposal facility depends on the environmental conditions in that facility (see Chap- ter 6). Waste form acceptability also depends on the performance require- ments for that disposal facility (Ebert, 2008; see Chapter 8). The acceptability of a waste form material for disposal is determined through predictive modeling studies (Chapter 7). These models use vari- ous thermodynamic and kinetic approaches, but most models used in the United States are based on irreversible thermodynamic (or steady state) transition state theory (TST). The information derived from laboratory tests can be used to parameterize these predictive models (e.g., Grambow, 1985; Grambow and Muller, 2001).2 In some cases, it may be necessary to accelerate releases from a waste form material to obtain the necessary infor- mation. This can be done by altering the parameters of a laboratory test, for example, surface area (SA), time (t), temperature (T), or a combination such as (SA) × (t). This is a useful approach so long as the alterations do not change the release mechanisms. The testing of a waste form material in the laboratory can be related to acceptable performance of that material in a disposal facility by the fol- lowing linking relationships (Ebert 2008; Plodinec and Ramsey, 1994):3 process control ↔ composition control ↔ release rate control4 ↔ performance control ↔ acceptable performance (5.1) These linking relationships provide a logical technical approach for identifying an acceptable range of processing and composition controls based on the range of waste form release rates and level of disposal system performance deemed acceptable by regulators. In other words, these rela- tionships provide a technical basis for identifying an acceptable waste form release rate under particular test conditions because those test conditions can be related to performance. 2 The tests described in this chapter provide the fundamental information shown at the base of the Performance Assessment Pyramid (see Figure 7.2) that enables such modeling. 3 Other approaches could be used, depending on the process for producing the waste form. For example, one could combine careful control of processing conditions with a frequent sampling procedure to ensure that the proper product has been produced. 4 Dissolution rate control is achieved by modifying the composition and or waste-loading of the waste form.

OCR for page 119
123 WASTE FORM TESTING For high-level waste (HLW) glass (i.e., alkali borosilicate glass) this linking relationship was established in a stepwise fashion: • Develop an acceptable waste form durability specification based on HLW performance modeling. As discussed in Chapter 8, acceptable fractional release rates for waste forms in a generic geologic reposi- tory were determined to be between 10–4 to 10–6 parts per year5 (Crandall, 1983). Because early versions of 10 CFR Part 60.1136 specified fractional release rates of 10–5 parts per year, which was in the middle of the range determined by HLW performance model- ing, this rate was adopted as the waste form specification. • Select a waste form material that had the potential to meet this specification. As discussed in Chapter 8, borosilicate glass was selected as a waste form material for several reasons, including its potential ability to meet this performance specification. • Develop an understanding of borosilicate glass durability mecha- nisms from a combination of ASTM test protocols (ASTM C1220, ASTM C1285, ASTM C1662), which were then a suite of tests under development by the Materials Characterization Center (MCC) (see Appendix 5.A for a history of test development). These test protocols are described in Appendix 5.C. • Develop a glass standard, the Environmental Assessment (EA) glass, which bounded the upper release rate found to be accept- able from HLW performance modeling and 10 CFR Part 60.113.7 • Generate a substantial database for modeling the maximum radio- active release rate(s), which happens to be for technetium-99, iodine-129, and cesium-135,8 by the release of non-radioactive species such as sodium, lithium, and boron, which release at the same rate (i.e., congruently; see Box 5.2). • Develop a standard test for ensuring that every glass produced has a release rate less than that of the EA glass based on sodium, lithium, and boron, which in turn ensures performance control and acceptable performance. • Using this standard test, continue to periodically verify that the durability of the production glass meets performance specifications. 5 That is, the waste form would take 104 to 106 years to completely dissolve. 6 Waste Package Performance Objective; see Chapter 8. 7 This acceptable release rate was based on a bounding calculation for a generic repository. The actual dissolution rate of a waste form material after emplacement in a disposal facility will depend on the specific geochemical and hydrological characteristics of that facility. 8 Technetium-99, iodine-129, and cesium-135 are not solubility limited; consequently, these radionuclides are released at maximum forward (initial) rates of dissolution.

OCR for page 119
124 WASTE FORMS TECHNOLOGY AND PERFORMANCE BOX 5.2 Congruent vs. Incongruent Dissolution The term congruency describes the dissolution behavior of atomic species (including radioactive species) in a waste form material as that material reacts with a solution. If species dissolve in proportion to their presence in a waste form material (i.e., in stoichiometric proportions), then dissolution is said to be congruent. In such cases, the rate of release of species from the waste form is proportional to both the dissolution rate of the waste form and the relative abundance of those species in the waste form. For materials that exhibit this behavior, for example borosilicate glass in a high pH under-saturated solution, the dissolution behavior of non-radioactive species such as sodium, lithium, and boron can be conveniently used to monitor the releases of radionuclides such as technetium-99, iodine-129, and cesium-135. Decades of research have provided the basis for this relation- ship to be used for HLW borosilicate glass (Bates et al., 1983; Bazan et al., 1987; Bibler and Bates, 1990; Bibler and Jurgensen, 1988; Bradley et al., 1979; Ebert et al., 1996; Fillet et al., 1985; McGrail, 1986; Ojovan et al., 2006; Vernaz and Godon, 1992). If some species in the waste form material dissolve preferentially to thers, o then dissolution is said to be incongruent. Incongruent dissolution is often d iffusion-controlled and can be surface reaction affinity-limited under conditions of near saturation or mass transport-controlled. Preferential phase dissolution, ion- exchange reactions, grain-boundary dissolution, and dissolution-reaction product formation (surface crystallization and recrystallization) are among the more likely mechanisms of incongruent dissolution. Precipitation of a secondary phase or phases can also lead to incongruent dissolution. Apparent incongruent dissolution can occur in complex monophase or poly- phase crystalline ceramic waste forms. For example, a multiphase ceramic waste form may contain sodium in more than one phase, whereas species such as technetium-99 are only sequestered in one of the sodium-containing phases. In this case, each phase undergoes congruent dissolution, but technetium-99 and sodium will not be released into solution at the same rate. This approach was the basis for qualifying HLW glass from West Valley and the Savannah River Site in the Yucca Mountain Total System Performance Assessment–License Application (TSPA-LA) (Ebert, 2000). It was also the approach used in the Hanford performance evaluation for low-activity waste (LAW) glass intended for shallow-land burial (Mann et al., 2001) and to qualify glass-bonded sodalite for disposal in a deep geological repository (Ebert, 2005). A similar approach was also taken by Ebert (2005) to qualify a metallic waste form. The HLW EA glass (Jantzen et al., 1993, 1994) standard does not necessarily apply to other types of glasses. For example, another glass

OCR for page 119
125 WASTE FORM TESTING standard, the LAW Reference Material (LRM) glass standard,9 was devel- oped for Hanford’s Immobilized LAW (ILAW) glass. The LRM contained the glass components anticipated to be present in the Hanford low-activity waste streams as well as those that may be added to facilitate vitrification or improve the durability of the ILAW waste products (Ebert and Wolf, 1999). Extensive testing was again used to demonstrate that the most sol- uble radionuclides (i.e., those that are not solubility limited) were released congruently (Box 5.2) to sodium, lithium, and boron in the glass to qualify it for near-surface disposal. 5.2 TEST SELECTION A suite of standard laboratory tests have been developed (Appendix 5.C) to measure the release behavior of waste form materials. The selection of a particular test for a particular waste form material depends on that mate- rial’s release mechanism (Table 5.1). Standard tests established for use on materials that release by one mechanism, such as glass that preferentially releases its constituents by reaction affinity-control under non-saturated conditions, cannot necessarily be applied to materials that release by a dif- ferent mechanism, such as cement that releases constituents by diffusion (e.g., Ojovan and Lee, 2005). Similarly, one cannot apply standard tests for borosilicate glasses to non-borosilicate glasses, because it is not known whether constituents in the latter material release congruently (Box 5.2) by the same mechanism(s). In these cases, new standard tests need to be developed, or existing standard tests need to be qualified, once the release mechanisms for a new material are determined. The recent determination of the release mechanisms for silicate glasses and minerals provide a good illustration of this point. The rate-limiting step in silica-water reactions in a glass or mineral is breakage of the structural Si–O bonds (Oelkers, 2001; Oelkers et al., 1994; Rimstidt and Barnes, 1980). Oelkers (2001) has shown that the release mechanisms for single- phase minerals and glasses are similar. Thus, modeling of the dissolution of glass has paralleled the modeling of mineral-solution dissolution. Kinetic treatments have systematized the effects of pH, temperature, saturation state, ionic strength of the leachant, and inhibition on the overall release rate by developing models that treat each effect individually (Lasaga and Luttge, 2004). The kinetic effects of saturation state as a function of pH, temperature, and ionic strength have primarily been handled by the appli- cation of combined thermodynamic and kinetic TST models and the free energy dependence of basic irreversible dissolution reactions (Aagaard and 9 This glass was originally developed as a standard for test method responses and later became a standard for glass durability (see Ebert and Wolf, 1999; Wolf et al., 1998).

OCR for page 119
  126 WASTE FORMS TECHNOLOGY AND PERFORMANCE TABLE 5.1 Summary of Waste Form Durability Response and Tests Waste Form Class Retention Mechanism Single-Phase Glasses Chemical incorporation Constituentsb 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). Glass-Ceramic Material Chemical incorporation Constituents are present in the glass matrix and benign crystals such as spinels (Cr, Ni, and Fe species) are allowed to crystallize ( ). These crystals do not contain radionuclides but may contain hazardous constituents (e.g., Cr, Ni). Glass-Ceramic Material Chemical incorporation and encapsulation Constituents are present in the glass matrix and in the crystalline phases. Example shows Cs in the glass and in a secondary phase ( ). Secondary phase may be more soluble than glass (e.g., (Na,Cs)2SO4) or more durable than glass (e.g., pollucite (Cs,Na)2Al2Si4O12). Single-Phase Oxides/Minerals/Metals Chemical incorporation Consists of only one main crystalline phase, which contains the same radionuclide(s). May be granular or monolithic.

OCR for page 119
127 WASTE FORM TESTING Durability Appropriate Durability Test/ Standardsa Graphical Representation Behavior single phase ASTM C1220 EA, ARG-1, source – ASTM C1285 LRM, work for homogeneous ASTM C1662 borosilicate-based glass ASTM C1663 GCMs; testing and PUF standards must be developed for non-borosilicates single phase ASTM C1220 EA, ARG-1, Table 5.1-SinglePhaseGlasses.eps source— ASTM C1285 LRM, work for homogeneous ASTM C1662 borosilicate based glass as long as ASTM C1663 GCMs; testing and crystalline phases PUF standards must do not sequester be developed for constituents non-borosilicates multiphase ASTM C1220 EA, ARG-1, Table 5.1-GlassCeramicMaterial.eps source—glass ASTM C1285 LRM, work for and multiple ASTM C1662 borosilicate-based crystalline ASTM C1663 GCMs; testing and phases and grain PUF standards must boundaries be developed for non-borosilicates multiphase ASTM C1220 Testing and Table 5.1-GlassCeramicMaterial2.eps source—single ASTM C1285 standards must be crystalline ASTM C1662 developed phase and grain ASTM C1663 boundaries PUF continued Table 5.1-SinglePhaseOxides.eps

OCR for page 119
128 WASTE FORMS TECHNOLOGY AND PERFORMANCE TABLE 5.1 Continued Waste Form Class Retention Mechanism   Multiphase Oxides/Minerals/Metals Chemical incorporation Individual phases contain one or multiple constituents (e.g., solid solution indicated between UO2 and ThO2). Some phases do not incorporate any constituents (gray shading). May be granular or monolithic. Multiphase Granular Oxides/Minerals/Metals Chemical incorporation and encapsulation Granular waste forms must be monolithed for disposal if not containerized. The monolithing agent does not incorporate constituents (gray shading). Also known as composite waste forms. Cementation/Hydroceramics Encapsulation T Geopolymers Hydrated phases incorporate constituents weakly 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. NOTES: EA = Environment Assessment Glass; ARG-1 = Analytical Reference Glass-1; LRM = Low-Activity Waste Reference Material. a Standards are only appropriate if mechanisms and radionuclide releases are shown to be the same. The tests are described in Appendix 5.C. b Can include both radioactive and chemical constituents. Key: Cs U Tc Pu

OCR for page 119
129 WASTE FORM TESTING Durability Appropriate Durability Test/ Standardsa Graphical Representation Behavior multiphase ASTM C1220 Testing and source—multiple ASTM C1285 standards must be crystalline ASTM C1662 developed phases and grain ASTM C1663 boundaries PUF multiphase ASTM C1220 Testing and Table 5.1-MultiphaseOxides.eps source—multiple ASTM C1285 standards must be crystalline phases ASTM C1662 developed but binder and ASTM C1663 grain boundaries PUF contain no constituents ASTM C1308 or ANSI 16.1 or EPA 1315 multiphase ASTM C1308 Radionuclides able 5.1-MultiphasegranulatedOxides.eps source—multiple or or simulants must crystalline phases ANSI 16.1 or be measured or a but phases EPA 1315 standard developed encapsulate the constituents which exist primarily on the grain boundaries Table 5.1-CementationHydroceramics.eps

OCR for page 119
142 WASTE FORMS TECHNOLOGY AND PERFORMANCE steady-state concentrations of radionuclides can provide an important con- firmation to the application thermodynamic databases. This raises the importance of continuing the development of advanced state-of-the-art solids and solutions analytical techniques in support of accelerated waste form testing (e.g., Pierce, 2008). Data from PCT-B tests may form part of the larger body of data that are necessary for long-term prediction of waste form behavior (see Figure 5.1). PCT-B tests are useful for generating concentrated solutions to study chemi- cal affinity effects on the dissolution rate. Tests at high temperatures and high glass/solution mass ratios can be used to promote the formation of alteration phases to (1) identify the kinetically favored alteration phases, (2) determine their propensity to sequester radionuclides, and (3) evaluate the effect of their formation on waste form dissolution rate. This informa- tion can be used to support the development of waste-form alteration models that can be coupled with relevant aqueous transport models to predict the release rate of radionuclides over the very long time periods pertinent to the operation of an HLW repository. As noted previously, equilibrium or steady-state conditions may be achieved under the very extended service periods relevant to geological disposal. The same build-up of dissolved species that leads to a reduction in borosilicate dissolution rate also leads to saturation of the groundwater, with potential precipitation of both stable and radioactive dissolved species. At this point, the initial control of radionuclide release by reaction kinetics of the waste form would be replaced by solubility limits to radionuclide con- centrations imposed by the initial crystalline waste form matrix, or by for- mation of new alteration products. The extended capabilities and flexibility of the PCT-B tests are intended to establish the various processes and asso- ciated data that control the release of radionuclides from disposal systems as waste forms react with groundwater over thousands of years and more. ASTM C1662 (MCC-4) This durability test method, known as the Single-Pass Flow-Through (SPFT) Test Method, is used for the measurement of glass dissolution rates. This test is most frequently used on homogeneous glasses, including nuclear waste glasses, in various test solutions at temperatures less than 100°C. The test procedure allows for inhomogeneous glasses (i.e., those that are phase separated or crystallized) to be studied provided the test response from each phase can be determined. The SPFT test is best suited for use with crushed materials, but tests can be conducted with monolithic specimens (Tole, 1982; Tole et al., 1986). The SPFT test has been used for decades by geologists to measure the dissolution of minerals. It is commonly used for single-phase crystal-

OCR for page 119
143 WASTE FORM TESTING line ceramics but has also been used for multiphase mineral waste forms (Icenhower et al., 2003; Jantzen et al., 2007; McGrail et al., 2003a,b; Zhao et al., 2000). Data interpretation is more complex with multiphase mineral waste forms because there are different source terms coming from the dif- ferent mineral phases, unless comparisons can be made to the dissolution of single-phase natural analogue minerals and/or single-phase pure standards that have been tested for comparison. However the SPFT test is the most informative for characterization of a material’s leaching parameters and is recommended for determining the long-term dissolution of glass (Strachan, 2001). SPFT tests may be conducted under conditions in which the effects from dissolved species on the dissolution rate are minimized to measure the forward dissolution rate at specific values of temperature and pH, or to measure the dependence of the dissolution rate on the concentrations of various solute species. This test can be used to characterize various aspects of corrosion behavior that can be utilized in a mechanistic model for calcu- lating long-term behavior of a nuclear waste glass. Many of the parameters determined from this test, such as the activation energy of dissolution and the reaction progress, are used in the TST and irreversible intrinsic models developed for mineral dissolution (Helgeson et al., 1984; Oelkers, 2001; Oelkers et al., 1994). The composition of the leachant solution can be controlled precisely, and dissolution rates can be measured fairly precisely (see Figure 5.C.2). The effects of the solution flow rate and sample surface area are taken into account when determining the dissolution rate using the rate equation for glass/mineral dissolution. The test method is appropriate for other materi- als that dissolve by the same mechanism, such as aluminosilicate minerals (Ebert, 2008). The test can be used to measure effects of various leachant components when waste solution volume is not a limitation (e.g., with non- radioactive materials). The reacted sample recovered from a test may be examined with sur- face analytical techniques, such as scanning electron microscopy, to further characterize corrosion behavior. Such examinations may provide evidence whether the waste form is dissolving stoichiometrically or if particular leached layers and secondary phases were formed on the specimen surface. These occurrences may impact the accuracy of the glass dissolution rate that is measured using this method. ASTM C1663 The vapor hydration test (VHT) (Ebert et al., 1991) is a static test in which a monolithic specimen is suspended in a sealed vessel with a small amount of water. When heated, the vapor phase becomes saturated, and a

OCR for page 119
144 WASTE FORMS TECHNOLOGY AND PERFORMANCE FIGURE 5.C.2 Schematic of the SPFT test method. Figure 5.C.2.eps SOURCE: Eric Pierce,Oak Ridge National Laboratory. bitmap thin film of water condenses on the specimen. The amount of water in the vessel is carefully controlled so that no liquid remains. This is done to pre- vent solution from dripping off the specimen and establishing a reflux cycle and to maintain a static film of water on the specimen. Alteration phases formed on the reacted sample are analyzed, and thickness of the altered surface layer is measured on a cross-sectioned specimen. The VHT can be used to study the corrosion of glass and glass ceramic waste forms under conditions of high temperature and contact by water vapor or thin films of water. This method may serve as an accelerated test for some materials, because the high temperatures will accelerate thermally activated processes. A wide range of test temperatures have been reported in the literature, from 40°C (e.g., Ebert et al., 2005) to 300°C (e.g., Vienna

OCR for page 119
145 WASTE FORM TESTING et al., 2001). It should be noted that with increased test temperature comes the possibility of changing the corrosion rate-determining mechanism and the types of alteration phases formed from those that occur at lower tem- peratures such as in a particular disposal environment (Vienna et al., 2001). The VHT can be used as a screening test to determine the propen- sity of waste forms to alter and for relative comparisons in alteration rates between waste forms. This test provides useful information regarding the alteration phases that are formed,4 the disposition of radioactive and hazardous components, and the alteration kinetics under the specific test conditions. This information may be used in performance assessment (e.g., Mann et al., 2001). In a modification of the VHT, enough water is added to promote reflux- ing, and the solution is analyzed periodically to track the release of constitu- ents. This provides very high specimen surface/volume ratios in a test with a monolithic specimen. This modification is similar to the Soxhlet test, except that the sample itself is used to condense the water vapor and maintain an adhering layer of water. Thus, the modified VHT method serves as a simpli- fied flow-through test or Soxhlet test at elevated temperatures. PNNL PUF TEST The pressurized unsaturated flow (PUF) test (McGrail et al., 1997a) was developed at Pacific Northwest National Laboratory (PNNL) to simulate the flow of water/air mixtures in a hydrologically unsaturated environment. The test method is similar to the SPFT test in that the water/air mixture flow through a crushed sample and the effluent is collected periodically for analysis. The leachant can be pre-conditioned by placing other materials upstream of the sample, for example, to simulate interactions with geologic or engineering materials; interactions of released species can be simulated by placing other materials downstream of the sample. Reacted sample materials can be extracted and analyzed at the end of the test. This test can be used to directly incorporate materials interactions in the test and simulate integrated hydrologically unsaturated systems. Leachant composition is controlled prior to contacting the specimen and the solution chemistry resulting from corrosion can be tracked during the test. Altered specimen and alteration phases can be collected for analysis after testing. This test is appropriate for confirmation testing of waste form corrosion mechanism in an integrated environment, regardless of whether it is hydrologically saturated or unsaturated. The method is not well-suited 4The alteration phases that form in this test can be used as indicators of phases that might form under repository conditions.

OCR for page 119
146 WASTE FORMS TECHNOLOGY AND PERFORMANCE for tests with monolithic specimens because of uncertainties in the water flow path and specimen contact. The PUF test has not been standardized and is currently not conducted anywhere but PNNL, which has patented it (McGrail et al., 1999a). Some key uncertainties in the test are surface area of the crushed samples, pref- erential solution flow paths through sample, and possible modifications of the effluent prior to collection. The data resulting from several processes occurring in parallel or series can be difficult to relate to each specific pro- cess (McGrail et al., 1996, 1997a,b, 1999b). ASTM C1308 ASTM C1308 accelerated leach test (ALT) is a modification of the ANS/ ANSI 16.1 test method (see Figure 5.C.3) that can be used to (1) determine if the release of a component is controlled by diffusion and (2) determine the effective diffusion coefficient based on a model for diffusion from a finite cyl- inder. It is applicable to any matrix material that does not degrade or deform during the test, including cements and other monolithic waste forms. It is a semi-dynamic test in which a monolithic specimen of prescribed dimensions is immersed in a large volume of leachant in a sealed vessel for a relatively short interval. The leachate solution is periodically removed for analysis, and the sample is placed in fresh leachant to continue the test. The cumulative amounts of the species of interest released in successive test intervals are fitted with the diffusion equation for a finite cylinder. The test results can be used to qualitatively determine if the release of a component is controlled by diffusion alone, partitioned into a non-leachable fraction, or affected by solution saturation effects. Although evaluation of the diffusion coefficient requires use of a monolithic specimen having right cylinder geometry, the test method can be modified for use with crushed materials to determine (qualitatively) if releases are being controlled by diffusion. This test provides for the determination of an effective diffusion coeffi- cient using a mechanistic model. The method provides a procedure to deter- mine if release from small or irregular specimens is controlled by diffusion or matrix dissolution, even though the specimens cannot be modeled to determine a diffusion coefficient from the test data. Very large volumes of waste solution can result from testing. ANSI 16.1 This standard is similar to EPA Draft Method 1315 as well as ASTM C1308 (in fact, ANSI 16.1 preceded the ASTM C1308 standard). It pro- vides for less frequent replenishment of the leachate and the calculation of a leaching index for various radionuclides (see Figure 5.C.3). The test

OCR for page 119
147 WASTE FORM TESTING FIGURE 5.C.3 Schematic representation of the ANSI/ANS 16.1 test method. ASTM Figure 5.C.3.eps C1308 and EPA 1315 are similar but have more frequent replenishment frequencies. bitmap SOURCE: EPA, 2009. procedure is used to measure and index the release of radionuclides from waste forms as a result of dissolution in demineralized water for five days or longer. The results of this procedure do not apply to any specific envi- ronmental situation except through correlative studies of actual disposal site conditions. The test has by now become familiar to those working in the radioactive waste form development field. EPA 1315 This test protocol (EPA, 2009) is a relatively new procedure that is still undergoing round robin testing. It is designed to provide the mass transfer rates (release rates) of inorganic analytes contained in a monolithic or com- pacted granular material under diffusion-controlled release conditions as a function of dissolution time. Observed diffusivity and tortuosity may be estimated through analysis of the resulting dissolution test data. The test is suitable to a wide range of solid materials, which may be monolithic (e.g., cements, solidified wastes) or compacted granular materials (e.g., soils, sediments, stacked granular wastes) that behave as a monolith in that the predominant water flow is around the material and release is controlled by diffusion to the boundary. This test provides intrinsic material parameters for release of inorganic species under mass transfer-controlled dissolution conditions. It is intended as a means for obtaining a series of eluants, which may be used to estimate the diffusivity of constituents and physical retention parameter of the solid material under specified laboratory conditions. EPA 1315 is a characterization method and does not utilize solu- tions considered to be representative of field conditions. This method is

OCR for page 119
148 WASTE FORMS TECHNOLOGY AND PERFORMANCE similar in structure and use to predecessor methods ANSI/ANS 16.1 and ASTM C1308. However, this method differs from previous methods in that: (1) leaching intervals are modified to improve quality control, (2) sample preparation accounts for mass transfer from compacted granular samples, and (3) mass transfer may be interpreted by more complex release models that account for physical retention of the porous medium and chemical retention at the pore wall through geochemical speciation modeling.

OCR for page 119
149 WASTE FORM TESTING REFERENCES Aagaard, P. and H. C. Helgeson. 1982. “Thermodynamic and Kinetic Constraints on Reaction Rates Among Minerals and Aqueous Solutions, I. Theoretical Considerations,” Amer. J. Sci. 282, 237-285. ANDRA [National Radioactive Waste Management Agency]. 2005. Dossier 2005 Argile. Tome: Safety Evaluation of a Geological Repository, ANDRA Report Series, Châtenay- Malabry, France. Apted, M. J. 1982. “Overview of Hydrothermal Testing of Waste Package Barrier Materials at the Basalt Waste Isolation Project,” In Material Characterization Center Workshop on the Leaching Mechanisms of Nuclear Waste Forms, May 19-21, 1982, PNL-4382/ UC-70, Pacific Northwest Laboratory, Richland, Wash. Bates, J. K., D. J. Lam, and M. J. Steindler. 1983. “Extended Leach Studies of Actinide-Doped SRL 131 Glass,” In Scientific Basis for Nuclear Waste Management VI, D. G. Brookins (Ed.), North-Holland, New York, 183-190. Bazan, F., J. Rego, and R. D. Aines. 1987. “Leaching of Actinide-doped Nuclear Waste Glass in a Tuff-Dominated System,” In Scientific Basis for Nuclear Waste Management X, J. K. Bates and W. B. Seefeldt (Eds.), Materials Research Society, Pittsburgh, Penn., 447-458. Bibler, N. E. and A. R. Jurgensen. 1988. “Leaching Tc-99 from SRP Glass in Simulated Tuff and Salt Groundwaters,” In Scientific Basis for Nuclear Waste Management XI, M. J. Apted and R. E. Westerman (Eds.), Materials Research Society, Pittsburgh, Penn., 585-593. Bibler, N. E. and J. K. Bates. 1990. “Product Consistency Leach Tests of Savannah River Site Radioactive Waste Glasses,” In Scientific Basis for Nuclear Waste Management XIII, V. M. Oversby and P. W. Brown (Eds.), Materials Research Society, Pittsburgh, Penn., 327-338. Bradley, D. J., C. O. Harvey, and R. P. Turcotte. 1979. Leaching of Actinides and Technetium from Simulated High-Level Waste Glass, PNL-3152, Pacific Northwest Laboratory, Richland, Wash. Bruno, J. and A. Sandino. 1987. Radionuclide Co-precipitation, SKB-TR-87-23, Swedish Spent Fuel and Nuclear Waste Management Co., Stockholm, Sweden Crandall, J. L. 1983. “High Level Waste Immobilization,” In The Treatment and Handling of Radioactive Wastes, A. G. Blasewitz, J. M. Davis, and M. R. Smith (Eds.), Battelle Press and Springer-Verlag, 178-183. Day, D. E., X. Yu, G. J. Long, and R. K. Brow. 1997. “Properties and Structure of Sodium-iron Phosphate Glasses,” J. Non-Cryst. Solids 215(1), 21-31. DOE [U.S. Department of Energy]. 2008. Yucca Mountain Repository License Application: Safety Analysis Report, DOE/RW-0573, Rev 0., Office of Civilian Radioactive Waste Management, Las Vegas, Nev. Ebert, W. L. 2000. Defense High Level Waste Glass Degradation, Office of Civilian Radio- active Waste Management Analysis/Model, ANL-EBS-MD-000016, Rev. 0 ICN01. (December). Ebert, W. L. 2005. Testing to Evaluate the Suitability of Waste Forms Developed for Electro- metallurgically-Treated Spent Sodium-Bonded Nuclear Fuel for Disposal in the Yucca Mountain Repository, ANL-05/43, Argonne National Laboratory, Argonne, Ill. Ebert, W. L. 2008. Testing Protocols to Support Waste Form Development, Production, and Acceptance, GNEP-WAST-WAST-AI-RT-2008-000302, U.S. Department of Energy, Office of Nuclear Energy, Washington, D.C. (September). Ebert, W. L., J. K. Bates, and W. L. Bourcier. 1991. “The Hydration of Borosilicate Waste Glass in Liquid Water and Steam at 200°C,” Waste Manage.11, 205-221.

OCR for page 119
150 WASTE FORMS TECHNOLOGY AND PERFORMANCE Ebert, W. L., S. F. Wolf, and J. K. Bates. 1996. “The Release of Technetium from Defense Waste Processing Facility Glasses,” In Scientific Basis for Nuclear Waste Management XIX, W. M. Murphy and D. A. Knecht (Ed.), Materials Research Society, Pittsburgh, Penn., 221-227. Ebert, W. L. and S. F. Wolf. 1999. Round-Robin Testing of a Reference Glass for Low-Activity Waste Forms, ANL-99/22, Argonne National Laboratory, Argonne, Ill. EPA [U.S. Environmental Protection Agency]. 2009. Mass Transfer Rates of Constituents in Monolith or Compacted Granular Materials Using a Semi-Dynamic Tank Leaching Test, Draft Method 1315, Available at http://www.epa.gov/osw/hazard/testmethods/sw846/ index.htm. Fillet, S., J. Nogues, E. Vernaz, and N. Jacquet-Francillon. 1985. “Leaching of Actinides from the French LWR Reference Glass,” In Scientific Basis for Nuclear Waste Management IX, L. O. Werme (Ed.), Materials Research Society, Pittsburgh, Penn., 211-218. Grambow, B. 1985. “A General Rate Equation for Nuclear Waste Glass Corrosion,” Mat. Res. Soc. Symp. Proc. 44, 15-27. Grambow, B. and R. Muller. 2001. “First-order Dissolution Rate Law and the Role of Surface Layers in Glass Performance Assessment,” J. Nucl. Mat. 298, 112-124. Helgeson, H. C., W. M. Murphy, and P. Aagaard. 1984. “Thermodynamic and Kinetic Con- straints on Reaction Rates Among Minerals and Aqueous Solutions, II. Rate Constants, Effective Surface Area, and the Hydrolysis of Feldspar,” Geochim. et Cosmochim. Acta 48, 2405-2432. Icenhower, J. P., D. M. Strachan, M. M. Lindberg, E. A. Rodriguez, and J. L. Steele. 2003. Dissolution Kinetics of Titanate-Based Ceramic Waste Forms: Results from Single-Pass Flow Tests on Radiation Damaged Specimens, PNNL-14252, Pacific Northwest National Laboratory, Richland, Wash. (May). Jantzen, C. M., N. E. Bibler, D. C. Beam, and M. A. Pickett. 1993. Characterization of the Defense Waste Processing Facility (DWPF) Environmental Assessment (EA) Glass Standard Reference Material, WSRC-TR-92-346, Rev. 1, Westinghouse Savannah River Company, Aiken, S.C. (February). Jantzen, C. M., N. E. Bibler, D. C. Beam, and M. A. Pickett. 1994. “Development and Char- acterization of the Defense Waste Processing Facility (DWPF) Environmental Assessment (EA) Glass Standard Reference Material,” In Environmental and Waste Management Issues in the Ceramic Industry, Ceram. Trans. 39, 313-322. Jantzen, C. M., T. H. Lorier, J. M. Pareizs, and J. C. Marra. 2007. “Fluidized Bed Steam Reformed (FBSR) Mineral Waste Forms: Characterization and Durability Testing,” In Scientific Basis for Nuclear Waste Management XXX, D. Dunn (Ed.), 379-386. Lasaga, A. C. 1984. “Chemical Kinetics of Water-Rock Interactions,” J. Geophys. Res. B6, 4009-4025. Lasaga, A. C. and A. Luttge. 2004. “Mineralogical Approaches to Fundamental Crystal Dis- solution Kinetics,” Amer. Mineral. 89, 527-540. Mann, F. M., R. J. Puigh II, S. H. Finfrock, E. J. Freeman, R. Khaleel, D. H. Bacon, M. P. Bergeron, B. P. McGrail, S. K. Wurstner, K. Burgard, W. R. Root, and P. E. LaMont. 2001. Hanford Immobilized Low-Activity Tank Waste Performance Assessment: 2001 Version, DOE/ORP-2000-24, Rev. 0, U.S. Department of Energy, Office of River Protec- tion, Richland, Wash. McGrail, B. P. 1986. “Waste Package Component Interactions with Savannah River Defense Waste Glass in a Low-Magnesium Salt Brine,” Nucl. Tech. 168-186.

OCR for page 119
151 WASTE FORM TESTING McGrail, B. P., C. W. Lindenmeier, P. F. Martin, and G. W. Gee. 1996. “The Pressurized Unsaturated Flow (PUF) Test: A New Method for Engineered-Barrier Materials Evalu- ation,” In Environmental Issues and Waste Management Technologies in the Ceramic and Nuclear Industries II, 72, V. Jain and D. K. Peeler (Eds.), American Ceramic Society, Westerville, Ohio, 317-329. McGrail, B. P., P. F. Martin, and C. W. Lindenmeier. 1997a. “Accelerated Testing of Waste Forms Using a Novel Pressurized Unsaturated Flow (PUF) Method,” Mat. Res. Soc. Symp. Proc. 465, 253-260. McGrail, B. P., W. L. Ebert, A. J. Bakel, and D. K. Peeler. 1997b. “Measurement of Kinetic Rate Law Parameters on a Na-Ca-Al Borosilicate Glass for Low-Activity Waste,” J. Nucl. Mat. 249, 175-189. McGrail, B. P., P. F. Martin, and C. W. Lindenmeier. 1999a. “Method and Apparatus for Measuring Coupled Flow, Transport, and Reaction Processes Under Liquid Unsaturated Flow Conditions,” Patent No. 5974859, Battelle Memorial Institute, Columbus, Ohio. McGrail, B. P., C. W. Lindenmeier, and P. F. Martin. 1999b. “Characterization of Pore Structure and Hydraulic Property Alteration in Pressurized Unsaturated Flow Tests,” In Scientific Basis for Nuclear Waste Management XXII, D. J. Wronkiewicz and J. H. Lee (Eds.), Material Research Society, Pittsburgh, Penn., 421-428. McGrail, B. P., H. T. Schaef, P. F. Martin, D. H. Bacon, E. A. Rodriguez, D. E. McCready, A. N. Primak, and R. D. Orr. 2003a. Initial Suitability Evaluation of Steam-Reformed Low Activity Waste for Direct Land Disposal, PNWD-3288, Battelle, Pacific Northwest Division, Richland, Wash. McGrail, B. P., E. M. Pierce, H. T. Schaef, E. A. Rodriguez, J. L. Steele, A. T. Owen, and D. M. Wellman. 2003b. Laboratory Testing of Bulk Vitrified and Steam-Reformed Low- Activity Forms to Support a Preliminary Assessment for an Integrated Disposal Facility, PNNL-14414, Pacific Northwest National Laboratory, Richland, Wash. Mendel, J. E. 1983. “Waste Glasses–Requirements and Characteristics,” In The Treatment and Handling of Radioactive Wastes, A. G. Blasewitz, J. M. Davis, and M. R. Smith (Eds.), Battelle Press, Columbus, Ohio, 178-183. Nagra [National Cooperative for the Disposal of Radioactive Waste]. 2002. Demonstra- tion of Disposal Feasibility for Spent Fuel, Vitrified High-Level Waste and Long-Lived Intermediate-Level Waste, TR-02-05, Nagra, Baden, Switzerland. NRC [National Research Council]. 1996. The Waste Isolation Pilot Plant: A Potential Solu- tion for the Disposal of Transuranic Waste, National Academy Press, Washington, D.C. NRC. 2005. Tank Wastes Planned for On-Site Disposal at Three Department of Energy Sites: The Savannah River Site—Interim Report, National Academies Press, Washington, D.C. Oelkers, E. H. 2001. “General Kinetic Description of Multioxide Silicate Mineral and Glass Dissolution,” Geochim. Cosmochim. Acta, 65(21), 3703-3719. Oelkers, E. H., J. Schott, and J. L. Devidal. 1994. “The Effect of Aluminum, pH and Chemical Affinity on the Rates of Aluminosilicate Dissolution Reactions,” Geochim. Cosmochim. Acta 58, 2011-2024. Ojovan, M. I. and W. E. Lee. 2005. An Introduction to Nuclear Waste Immobilisation, Elsevier, Amsterdam. Ojovan, M. I., A. S. Pankov, and W. E. Lee. 2006. “The Ion Exchange Phase in Corrosion of Nuclear Waste Glasses,” J. Nucl. Mat. 358, 57-68. ONWI [Office of Nuclear Waste Isolation]. 1981. Interim Performance Specifications for Waste Forms for Geologic Isolation, NWTS-19 DRAFT, Office of Nuclear Waste Isola- tion, Columbus, Ohio (October).

OCR for page 119
152 WASTE FORMS TECHNOLOGY AND PERFORMANCE Pierce, E. M., B. P. McGrail, P. F. Martin, J. Marra, B. W. Arey, and K. N. Geiszler. 2007. “Accelerated Weathering of High-Level and Plutonium-Bearing Lanthanide Borosilicate Waste Glasses Under Hydraulically Unsaturated Conditions,” Appl. Geochem. 22(9), 1841-1859. Plodinec, M. J. and W. G. Ramsey. 1994. “Glass Consistency and Glass Performance,” Spectrum 94 Proceedings, WSRC-MS-94-00311, Available at http://www.osti.gov/bridge/ servlets/purl/10163781-GCJCOo/native/. Rimstidt, J. D. and H. Z. Barnes. 1990. “The Kinetics of Silica-Water Reactions,” Geochim. Comochim. Acta 44, 1683-1699. Steefel, C. I., D. J. DePaolo, and P. C. Lichtner. 2005. “Reactive Transport Modeling: An Es- sential Tool and a New Research Approach for the Earth Sciences,” Earth Planet. Sci. Lett. 240, 539-558. Strachan D. M. 2001. “Glass Dissolution: Testing and Modelling for Long-Term Behaviour,” J. Nucl. Mat. 298, 69-77. Tole, M. P. 1982. “Factors Controlling the Kinetics of Silicate-Water Interactions,” Unpub- lished PhD Thesis, The Pennsylvania State University, University Park, Penn. (March). Tole, M. P., A. C. Lasaga, C. Pantano, and W. B. White. 1986. “The Kinetics of Dissolution of Nepheline (NaAlSiO4),” Geochim. Cosmochim. Acta 50(3), 379-392. Vienna, J. D., P. Hrma, A. Kiricka, D. E. Smith, T. H. Lorier, I. A. Reamer, and R. L. Schulz. 2001. Hanford Immobilized LAW Product Acceptance Testing: Tanks Focus Area Re- sults, PNNL-13744, Pacific Northwest National Laboratory. Richland, Wash. Vernaz, E. Y. and N. Godon. 1992. “Leaching of Actinides from Nuclear Waste Glass: French Experience,” In Scientific Basis for Nuclear Waste Management XV, C. G. Sombret (Ed.), Materials Research Society, Pittsburgh, Penn., 37-48. Weber, W. J., R. C. Ewing, and Lu-Min Wang. 1994. “The Radiation-Induced Crystalline-to- Amorphous Transition in Zircon,” J. Mat. Res. 9(3), 688-698. Wolf, S. F., W. L. Ebert, J. S. Luo, and D. M. Strachan. 1998. A Data Base and a Standard Material for Use in Acceptance Testing of Low-Activity Waste Products, ANL-98/9, Argonne National Laboratory, Argonne, Ill. Zhao, P., S. Roberts, and W. Bourcier. 2000. Technical Progress Report on Single Pass Flow Through Tests of Ceramic Waste Forms for Plutonium Immobilization, UCRL- ID-143361, Rev. 1, Lawrence Livermore National Laboratory, Livermore, Calif.