Waste Forms Technology and Performance: Final Report (2011)

7

Waste Form Performance in Disposal Systems

The 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 provides a discussion of the modeling portion of this charge, including waste form performance in disposal systems and models for evaluating waste form performance. Waste form testing is discussed in Chapter 5.

For the purposes of the discussion in this chapter it is important to distinguish between a disposal facility and a disposal system. The term disposal facility (see Chapter 6, Figure 6.1) refers to physical infrastructure; it typically includes tunnels (in the case of deep disposal) or surface excavations (in the case of shallow disposal), the surrounding host rock, and engineered barriers, including the waste form if present. A disposal system, on the other hand, refers to both physical infrastructure and how the natural and engineered barriers in that infrastructure function to sequester radioactive and hazardous constituents. The ability of a disposal system to sequester these constituents is referred to as disposal system performance. The performance of a disposal system can be evaluated using performance assessment (see Section 7.2).

7.1 WASTE MANAGEMENT SYSTEMS

As discussed in Chapter 2, waste processing and waste form production are key activities in the Department of Energy, Office of Environmental

Management’s (DOE-EM’s) cleanup program and, indeed, in any integrated waste management system. A conceptual diagram showing the important steps in DOE-EM’s waste-management system is provided in Figure 7.1. There are interdependencies among the steps in this system; moreover, decisions at each step can affect waste management options and activities at subsequent steps.

The International Atomic Energy Agency (IAEA, 1995) addresses these interdependencies explicitly in Principle 8. Radioactive Waste Generation and Management Interdependencies, which states:¹

In the context of the present report, this principle suggests the need to consider waste form development and selection in the context of disposal system performance. The principle also explicitly recognizes that, although there are many steps and activities that can optimize the safety, capacity, schedule, and cost of a waste management system, all paths eventually lead to final disposal.

7.2 DISPOSAL SYSTEM PERFORMANCE

The National Research Council (NRC) has published numerous reports bearing on the performance of disposal systems for spent nuclear fuel (SNF) and high-level radioactive waste (HLW) (e.g., NRC, 1995), transuranic (TRU) waste (NRC, 1996), and low-level radioactive waste (LLW) (e.g., NRC, 2005), as well as specifically on the performance of disposal systems (e.g., NRC, 1983, 1990, 2003). The NRC Committee on Technical Bases for Yucca Mountain Standards (NRC, 1995, p. 70) provided the following definition for disposal system performance, which, as noted previously, is usually referred to as performance assessment (PA):

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¹ This same principle to consider interdependencies in waste management operations and disposal is part of the Chapter 2, Article 4.iii of the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management, of which the United States is a signatory.

FIGURE 7.1 Schematic illustration of DOE-EM’s waste management system.

NOTE: WAC = waste acceptance criteria (see Chapter 8).

PA modeling has several useful applications in the design and licensing of disposal facilities, such as a repository for SNF/HLW; for example, PA can be used to:

• Demonstrate compliance with regulatory requirements, typically health-risk metrics such as dose rate to a critical group. Abstracted and simplified versions of PA can also be used to communicate with concerned stakeholders about disposal system performance. These applications are considered to be the conventional roles of PA.
• Identify system components and processes that strongly affect the isolation of radionuclides within disposal systems to coordinate and guide repository design, site characterization, and safety assessment activities.
• Evaluate the radiological safety of disposal systems in the larger context of costs, schedules, alternative options, and optimization of overall waste management policies.

This chapter focuses primarily on the first type of PA modeling: i.e., assessment and demonstration of compliance with regulatory guidelines. It describes a logical and systematic approach for carrying out PA modeling—an approach that is not always used in real-world applications.

When selecting a PA model, it is important to consider its fitness-for-purpose—that is, its suitability for the intended application. In many instances a “best estimate” analysis is warranted, especially with respect to establishing regulatory compliance. For exploratory purposes, conservative bounding analyses may be appropriate, although it must be cautioned that there is always a danger of compounding so many conservative assumptions and data values together that the resulting analysis becomes unrealistic and potentially misleading.

There are two important elements in a PA. First, the PA must consider the entire disposal system composed of multiple barriers and multiple containment processes. Modeling radionuclide releases and exposures requires consideration of the potential pathways by which radioactive constituents in the waste form could migrate through the disposal facility and eventually lead to future radiological exposures (health risks). Second, PA must consider the future behavior of these barriers (including the waste form) with respect to safety standards, such as those developed by the Environmental Protection Agency (see Chapter 8). The NRC (1995) recommended that a risk-based health standard should be applied as the appropriate metric for assessing the long-term safety of geological disposal for radioactive waste.

Repository programs typically employ a hierarchy of PA models to assess long-term safety, barrier design, and regulatory compliance of disposal systems containing radioactive waste. This hierarchy, referred to here as the PA Pyramid, is illustrated in Figure 7.2. The models of fundamental physicochemical processes affecting repository performance form the base of this pyramid. These models focus on processes such as chemical reactions among the waste form, other engineered barriers systems (EBS) in the disposal facility, and groundwater. Successively higher levels of PA models represent abstractions (with computational simplifications) of these underlying models. At intermediate levels in the PA Pyramid, the design and layout of EBS in the disposal facility are incorporated into the models. Models at the top of the pyramid represent abstractions with computational simplifications of the underlying models into a total system performance assessment, or TSPA (DOE, 2008; Whipple, 2006).

Uncertainties in assumptions, alternative conceptual models, and data are passed upward through the PA levels to ensure that all identified uncertainties are maintained at each level (e.g., DOE, 2008; Nagra, 2002). Likewise, sensitivity analyses made at an upper PA level can be used to identify specific risk-important factors or processes for which more detailed modeling and analysis at a lower PA level may be desirable. In other words, the PA Pyramid represents an iterative process for assessing the performance the waste form and its intended disposal system.

FIGURE 7.2 The PA pyramid showing the hierachical structure of PA models to evaluate the safety of disposal for radioactive waste. Uncertainties in assumptions, alternative conceptual models, and data are passed upward through the PA levels (upward arrow). Sensitivity analyses can be used to identify specific risk-important factors or processes for which more detailed modeling and analysis may be desirable (downward arrow).

7.3 MODELS FOR WASTE FORM PERFORMANCE IN DISPOSAL SYSTEMS

As noted in Chapter 6, the dominant potential pathway for radionuclide release from a disposal facility to the biosphere is via groundwater transport. Radioactive constituents may be released either as dissolved species or as suspended, radionuclide-bearing colloids. There is also the possibility of gaseous release of radionuclides that are volatile or form volatile compounds.² Because groundwater flow and radionuclide transport

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² A repository constructed in salt would likely have no aqueous, colloidal, or gaseous pathways for radionuclide release, unless there is some disruptive future human-based or natural event (NRC, 1996).

are primary concerns with respect to long-term performance of disposal systems for radioactive waste, the following conclusions from NRC (1990, pp. 14-15) are pertinent:

Transport processes link waste form dissolution (Chapter 5) to disposal system performance and safety. Transport of dissolved and colloidal species released by the dissolution of a waste form may be controlled by either advective flow of the groundwater, or, if the engineered or natural barriers surrounding the waste form have sufficiently low permeability (Chapter 6), by diffusion.

The following three subsections provide more detailed descriptions of the PA model hierarchy illustrated in Figure 7.2 and the key physical and chemical processes that affect disposal system performance.

7.3.1 PA Models

The NRC’s Waste Isolation System Panel (WISP) report (NRC, 1983) was the first multi-disciplinary study to integrate waste form dissolution and transport of dissolved radionuclides into a PA model for disposal systems. The WISP report applied well-understood, mass-transfer analytical models to elucidate the linkages between dissolution and transport in disposal systems. Independently, similar analytical mass-transfer models were being developed for international repository programs (e.g., KBS, 1983; Nagra, 1985; Neretnieks, 1978). The application of mass-transfer models to disposal system performance assessment is now routine for all types of radioactive waste (e.g., Andra, 2005; JNC, 2000; Nagra, 2002; SKB, 2006). Such simple mass-transfer models have, over time, been supported by more detailed and data-intensive physicochemical modeling (e.g.,

Steefel et al., 2005) in a manner consistent with the hierarchy of models shown in Figure 7.2.

7.3.1.1 Basic Release and Transport Processes that Control System Performance

To illustrate the linkage between waste-form dissolution and radionuclide transport, a simple but rigorous mass-transfer analytical model (NRC, 1983; Zavoshy et al., 1985) is cited here. The model is for a simplified geometry of a dissolving waste form surrounded by a diffusion barrier³ (i.e., buffer/backfill) and is based on well-validated, mass-transfer principles. A so-called flux ratio R is defined as (Zavoshy et al., 1985):

where

j₀ is the dissolution rate (i.e., chemical durability) of the waste form, normalized for the mass fraction of radionuclide i in the waste form⁴

r₀ is the radius of the waste form

e is the connected porosity of the buffer

D_e is the effective diffusion coefficient of dissolved radionuclide i in the buffer

C^SAT is the solubility limit for a solid phase incorporating radionuclide i.

For conditions where the flux ratio R is much greater than 1 (fast dissolution rate relative to diffusive transport flux), the long-term concentration of radionuclide i increases at the waste-form surface until the point that a solid phase containing radionuclide i precipitates; this precipitate sets a solubility-limited concentration (C^SAT) for radionuclide i at the waste form

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³ In all concepts for the disposal of SNF/HLW in saturated rock, a low-permeability buffer or backfill engineered barrier is placed around the SNF/HLW containers (e.g., NWTRB, 2009; Witherspoon and Bodvarsson, 2006). Such a buffer has several important safety functions, including promoting diffusive transport of all radionuclides released from the dissolution of waste forms. Low-permeability buffers also promote the filtration of any radionuclide-bearing colloids that might form from dissolution of the waste form (Nagra, 1994; SKB, 2006). For a repository located in unsaturated rock, buffer/backfills (so-called Richards barriers) based on the principle of capillary-breaking have been tested and built. These barriers are designed to perform the twin safety functions of assuring diffusive transport and colloid filtration (e.g., EPRI, 1996; Gee et al., 2002).

⁴Mass fraction is the mass of radionuclide i (i.e., waste loading of radionuclide i) divided by the mass of the entire waste form.

surface. This is called transport control, and the same basic process has long been recognized and applied to the interpretation of diagenetic mineral dissolution in nature (e.g., Berner, 1979). Conversely, when the flux ratio R is much lower than 1, the long-term concentration of radionuclide i at the waste form surface is basically controlled by the dissolution rate of the waste form. This is called surface-reaction control and also is well known from studies of natural diagenetic systems.

Note that this simple model does not depend on any specific waste form, dissolution rate mechanism, or model; it is general to any waste form for which a long-term dissolution rate (or release rate of radionuclides) can be defined. Similar mass-transfer expressions that link waste-form dissolution rate and advective transport have also been developed (Pigford and Chambré, 1987). These mass-transfer analytical models provide the critical linkage between waste form fabrication and geological disposal (Figure 7.1).

The important characteristics of waste forms with respect to long-term (10³-10⁶ years) performance and safety of a disposal system depend on a number of factors:⁵

• Type of waste form (see Chapter 3)
• Radionuclide inventory and waste loading of the waste form
• Environmental conditions in the near field of the disposal facility (see Chapter 6)
• Long-term dissolution rate of the waste form under those environmental conditions (see Chapter 5)
• Solubility limits of dose-contributing radioelements
• Rate of diffusive or advective aqueous transport of dissolved and colloidal radionuclides
• Presence of engineered barriers (e.g., clay buffer, Richards barrier)

For SNF/HLW repositories that include a low-permeability buffer surrounding the waste form, it is the solubility limits of the solid phases that incorporate the radionuclides that (i.e., C^SAT in Equation 7.1) are the dominant factors in limiting the long-term release rates of most radionuclides (e.g., Andra, 2005; DOE, 2008; JNC, 2000; Nagra, 2002). Such solubility limits are also considered to be controlling factors for potential releases via human intrusion from the WIPP site for disposal of defense TRU waste (DOE, 1995; NRC, 1996). Performance analyses of LLW disposal systems also typically apply these solubility limits as controls on radionu-

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⁵ The timeframe for regulatory compliance and the half-lives of key radionuclides present in the waste are also considerations; the dissolution rate of a waste form may limit radionuclide releases from disposal systems for an initial period before the onset of solubility limits imposed by precipitation of radionuclide-bearing solids.

clide releases (e.g., Andra, 2005; Nagra, 2002; NRC, 2005). Waste-form dissolution rates can, however, provide an important constraint on the release of highly soluble radionuclides, such as carbon-14, chlorine-36, and iodine-129, if such radionuclides are present in the waste form.

Application of PA models for disposal systems can place diverse factors such as waste-form dissolution rate, waste loading, and solubility limits of the solid phases containing radionuclides into a common system-level context for evaluation and optimization. Furthermore, such PA models can also provide guidance to future decisions on whether there is a safety-based reason for further development of advanced waste forms. The most notable application of system-level PA models to advanced waste form development relates to thresholds at which extremely low waste form dissolution rates would constrain (and simplify the calculation of) the performance of a given disposal concept.

NRC (1983, pp. 279-280) made a detailed analysis of the necessary fractional dissolution rate for waste forms performance to control the performance of disposal systems:

NRC (1983) further identified potential advantages of a low-solubility waste form with such exceptionally low fractional dissolution rates, including the following:

• Verification of the safety performance of the entire disposal system would depend largely on the laboratory measurements made under appropriate site-specific conditions (i.e., risk-based testing of waste forms),
• Release rates of an increasing number of radionuclides would become proportional to decreasing fractional dissolution rate,
• The need for estimating separate solubility limits would be greatly attenuated if not eliminated, and
• The number of sites that could serve as suitable repositories might increase.

A more recent analysis (SKB, 2006, Figure 10-44) of a deep geological repository for the disposal of spent nuclear fuel in granitic crystalline rock suggests the fractional dissolution-rate threshold for a waste form (in this case UO₂) might be as low as 10^–6/year for certain key radionuclides.

The exact threshold at which the waste-form dissolution rate controls the release performance of a disposal system (i.e., when the flux ratio R in Equation 7.1 becomes much less than 1) depends on a number of factors. In particular, waste loading and solubility limits, which typically vary among different radioelements over many orders of magnitude, influence this threshold value.

The sensitivity of factors affecting the threshold value at which waste-form fractional dissolution rate controls the release performance of a hypothetical disposal is illustrated in Table 7.1. This table applies Equation 7.1 to calculate flux ratios R for a several key, long-lived dose-contributing radioelements present in a standard HLW glass (Nagra, 2002) for reference waste loadings, solubility limits (C^SAT in Equation 7.1), and fractional dissolution rates.

For the reference fractional dissolution rate of 10^–5 parts per year, the releases of selenium-79, technetium-99, and neptunium-237 from the disposal system would be constrained by their respective solubility limits,

TABLE 7.1 Sensitivity of Calculated Flux Ratios Using Equation 7.1 for Radioelements with a Key Long-lived Radionuclide Present in a Reference HLW Borosilicate Glass

NOTES: Diffusional transport from the waste-form surface is assumed, with ε = 0.01, D_e = 3.15 × 10^–2 m²/year, and r₀ = 0.4 m (Zavoshy et al., 1985). R values much greater than 1 indicate release performance of the disposal system would be constrained by radioelement solubility, whereas R values much lower than 1 indicate release performance of the disposal system for that radionuclide would be constrained by waste form dissolution rate.

^a Waste loading for a Reference HLW Borosilicate Glass (McGinnes, 2002, Tables A.1-1 to A.1-4).

^b Reference Case radioelement solubilities for reducing disposal conditions (Nagra, 2002, Table A2.4).

^c The reported long-term dissolution rate of 5.5 × 10^–4 kg/m² year for the Reference HLW Borosilicate Glass is stated to correspond to a fractional dissolution rate of 10^–5 parts per year (Nagra, 2002, p. 144).

SOURCE: Nagra (2002).

whereas the release of highly soluble iodine-129 would be limited by the dissolution rate of the HLW glass. Even for a postulated fractional dissolution rate of 10^–7 parts per year, the releases of selenium-79, technetium-99, and neptnium-237 from the disposal system would still be constrained by their solubility limits. It would require a speculative fractional dissolution rate on the order of 10^–9 parts per year (i.e., the waste form would take 1 billion years to completely dissolve) for a waste form to control, and thereby lower, the release rates of these key radionuclides from the disposal system. This value is in basic agreement with the previous NRC (1983) estimate. The important point is that such sensitivity analyses provide a defensible basis by which to determine “how much better” an advanced waste form would have to perform to significantly enhance the safety of disposal systems compared to current HLW borosilicate glass, for example.

7.3.1.2 Integrated PA Models

The analytical models discussed above link the release and transport boxes shown in Figure 7.2. However, there are additional processes and barriers that affect the overall safety of disposal systems, including containment (i.e., barriers designed to delay contact between groundwater and waste forms), transport through the natural barrier (host rock) of the disposal facility, and finally the various pathways in which released radionuclides might migrate through the biosphere and lead to doses to humans. A system-level analysis is needed that incorporates all of the design aspects and properties of natural and engineered barriers that affect overall safety.

Numerical codes have been developed to allow more complete linkage among the models for the process boxes shown in the upper part of the PA Pyramid in Figure 7.2, such as the GoldSim code used in the recent license application for a SNF/HLW repository at Yucca Mountain (DOE, 2008) and the IMARC code (EPRI, 2009), which was also applied to the Yucca Mountain Site. Such codes provide great flexibility for evaluating uncertainties and sensitivities in model parameters; inclusion of alternative conceptual models for certain processes; detailed spatial expansion of important regions (compartments) of disposal facilities (especially the EBS); and relatively easy use of alternative data sets from pre-configured libraries.

Such top-level PA codes are now being used widely across DOE sites to provide a more accesible means for communicating about and addressing uncertainties and sensitivity about disposal system performance with nontechnical stakeholders. With the detailed models discussed in Sections 7.3.2 and 7.3.3, scientific and engineering understanding about the disposal system can be established. The top-level PA model is established with such fundamental understanding, while stylization for unverifiable assumptions, such as biosphere radionuclide pathway models, is introduced.

7.3.2 Intermediate Level Models

Significant insights on the performance of waste forms in disposal systems can be gained from the application of relatively simple performance assessment models described in the previous section. However, the abstractions in these models may not account for all of the important variables in the disposal system or changes in system conditions over time, especially during the initial period following facility closure. Consequently, there can be a need to develop engineering-type models that more fully incorporate the features of the facility design (e.g., EBS configuration, dimensions, and layout) and system conditions. Such models occupy the intermediate layer of the PA Pyramid in Figure 7.2 and bridge the fundamental process models at the bottom of the pyramid to the abstracted models at the top.

A facility for disposal of SNF/HLW will contain thousands of waste packages, usually in a two-dimensional array, each of which is surrounded by multiple engineered barriers. Some internal structure and heterogeneous radionuclide distribution will be present within each waste package. Consequently, the repository will display heterogeneity at different spatial scales. In conventional PA, radionuclide transport is modeled by reducing this heterogeneity to some extent (i.e., heterogeneity is homogenized). Packages are represented by several end-member types, and radionuclide transport in the repository is modeled without considering interferences from adjacent packages or the effects of the two-dimensional package-array configuration (Ahn et al., 2002).

The homogenization of spatial heterogeneity can obscure important processes that govern the performance of the disposal system, for example, the existence of advection-dominant flow paths. Radionuclide release from the near-field region to the far-field region is strongly influenced by the existence of these fast paths. The existence of fast paths can affect degradation of the engineered barriers, which in turn can affect fast-path geometry (Murakami and Ahn, 2008; Steefel et al., 2005).

Taking into account heterogeneity at all spatial scales requires a tremendous amount of computation. For instance, a relatively small-size simulated repository, containing fewer than 100 packages and millions of rock fractures in the near field, was the maximum size that could be simulated by the Earth Simulator supercomputer (Tsujimoto and Ahn, 2008). Compartmentalization is a logical modeling approach to improve computational efficiency because a repository contains thousands of packages surrounded by similar combinations of barriers. In a compartmentalization approach, modeling can be made at two levels: one at a local scale within a compart-

ment and the other at a repository scale connecting compartments. For waste form performance, local-scale modeling is more urgently needed.

7.3.3 Fundamental Process Models

Fundamental process models, which occupy the base of the PA Pyramid shown in Figure 7.2, integrate the knowledge obtained in the experiments and tests described in Chapter 5. These process models can be used to estimate performance at “local” (i.e., sub-facility) scales. Such models can be used to obtain best-estimates of waste form performance in particular disposal environments; to select suitable combinations of waste forms and engineered-barrier configurations; and to evaluate system performance using metrics other than dose, which can aid in optimizing facility designs.

More complex models for waste form durability need to account for waste form material properties (Chapter 3), disposal environment (Chapter 6), and interactions with other engineered and natural barriers in the disposal system (Chapter 6). The importance of such interactions is highlighted in a recent summary of the GLAMOR⁶ project (Van Iseghem et al., 2007, 2009). A specific focus of this project was to understand the long-term decrease in the rate of dissolution of glass waste forms (Figure 7.3), with two competing hypotheses considered:

1. The effect of silica concentrations in solution on the depression of the rate (the so-called “chemical affinity effect”⁷).
2. The role of surface layers that develop during the corrosion/dissolution process in limiting transport of reactive constituents to and from the primary glass phase, assuming that such layers do not spall off.

The rate of glass dissolution could be accelerated by placing it in proximity to a bentonite (clay) buffer or steel and iron canister corrosion products. Accelerated dissolution seems to be caused by the sorption and removal of glass reaction-products from solution, which if present would

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⁶ A Critical Evaluation of the Dissolution Mechanisms of High Level Nuclear Waste Glasses in Conditions of Relevance for Geological Disposal; see ftp://ftp.cordis.europa.eu/pub/fp6-euratom/docs/euradwaste04pro_pos9-van-iseghem_en.pdf.

⁷Chemical affinity is defined as the log (Q/K_sp), where Q is the ion activity product of dissolved species in solution and K_sp is equilibrium constant for the waste form (Lasaga, 1979). Borosilicate glass is thermodynamically unstable and cannot be re-precipitated from solution, so a proxy K_sp is derived for it.

FIGURE 7.3 Schematic representation of predominant mechanisms and resulting kinetics affecting the concentration of glass alteration elements silicon (Si), boron (B), and sodium (Na).

SOURCE: Van Iseghem et al. (2009).

slow the rate of dissolution due to the chemical affinity effect. To understand the underlying mechanism it was necessary to conduct a series of experimental studies supplemented with detailed microscopic characterizations of the evolving glass surface layers. Similar studies will be necessary for any candidate waste forms considered by DOE-EM because a mechanistic understanding of the controls on waste form dissolution provides the basis for understanding waste form performance at long time scales.

Evaluation of the performance of waste forms in disposal systems may be required for periods ranging up to 1 million years, depending on the pertinent regulations (see Chapter 8). Figure 7.4 provides an illustration of important processes that can occur in the near-field environment of a disposal facility for SNF/HLW over these time scales. The durability of a waste form depends, in addition to its own properties, on several environmental factors:

• solution composition and pH
• flow rate
• temperature

FIGURE 7.4 Simplified scheme of chemical processes in the near-field environment of a disposal facility.

SOURCE: Grambow et al. (2000), as modified by Horst Geckeis.

• redox conditions
• speciation in solution
• radiolysis
• interactions with corroded canisters and near-field geology
• formation and mobility of colloids

These individual factors often interact and are coupled in a repository environment. The dissolution of waste forms containing radioactive waste can be complex, particularly following the closure of a disposal facility when thermal, radiological, mechanical, hydrological, and chemical perturbations to the disposal system are highest. (This is a primary reason that radioactive wastes are typically placed in canisters with containment lifetimes of several thousands of years or more, which prevents groundwater contacting waste forms until these initial perturbations dissipate.)

Evaluating the complexity of disposal system performance can be accomplished using models that explicitly couple thermal, hydrological (transport), mechanical, and chemical processes in 3-dimensional representations of the barriers and spatial variation in properties. All of these processes occur against a backdrop of changing environmental conditions that may be externally imposed, including water infiltration rates, water chemistry, and decay of the thermal field. Local-scale mechanistic modeling

based on understanding of near-field processes can be made by coupling thermal (T), hydrological (H), chemical (C), radiological (R), and mechanical (M) processes for evolution and transport of materials in the near-field environment, including the waste form and its radioactive constituents. Estimates are typically restricted to pair-wise analyses such as T-H, T-C, T-M, or R-T, under the assumption that such sub-set analyses capture the full system behavior. Attempts to develop such models have already been made (e.g., Steefel et al., 2005) but much more work is needed. The development of such models could provide the scientific basis for best-estimates of waste form performance.

The challenge for using coupled models to evaluate waste form performance in disposal systems is the identification of key processes in the near-field environment, including:

• Rate-limiting steps. These could include the dissolution mechanism of the waste form; the formation and decomposition of radiolytically produced species in solution; initial surface sorption/desorption reactions; or the nucleation and precipitation of secondary phases.
• Reaction mechanisms: Even when rate-limiting steps have been identified, reaction mechanisms can change in response to changing environmental conditions. The full range of possible mechanisms and the conditions under which these mechanisms control waste form reactions must be evaluated.
• Radiolysis effects: Radiolysis at the waste form/solution interface can have an important effect on dissolution processes, particularly for redox-sensitive elements such as uranium and plutonium (see Chapter 6). The long-term effects of radiolysis, for example the formation of H₂O₂, have barely been explored, although in nature U(VI) phases that contain peroxyl groups have recently been discovered (Kubatko et al., 2003).

For coupled models to be realistic and useful, relevant chemical and physical processes must be represented. Relevant chemical processes include the following:

• Kinetically controlled dissolution of the waste form
• Nucleation and kinetic controls on the growth and sequence of metastable phases
• Solid-solution (co-precipitation, see Figure 7.4) models for substitution of minor components
• Kinetic and equilibrium sorption via different mechanisms (ion exchange, surface complexation)
• Aqueous complexation

• Oxidation-reduction state of the waste form as influenced by transport, local reactions, and radiolysis
• Composition and chemical concentration of pore water chemistry caused by evaporation

The relevant physical processes include the following:

• Heat transport as a result of convection, conduction, and radioactive decay, which provides the time-dependent temperature field affecting relative humidity, reaction kinetics, and thermodynamics
• Water flow under variably saturated conditions
• Diffusive and advective transport of solutes in the aqueous phase
• Gas phase transport via advection and diffusion, especially for the reactive gases O₂ and CO₂

The application of advanced reactive transport modeling of the near-field environment (Steefel et al., 2005) can aid in the overall safety assessment of disposal systems by reducing unwarranted conservatisms in more abstracted PA models and by enhancing the comprehensiveness and confidence in the PA.

7.4 DISCUSSION

The PA of waste forms containing radioactive waste can only be meaningfully accomplished within the context of disposal system PA, in which health-risk consequences are the appropriate basis for evaluations. As shown in Figure 7.2, there is typically a hierarchy of PA models employed in assessing any waste form/disposal system, with each level of PA models having the appropriate fitness for purpose, for example, design optimization, identification of risk-informed R&D needs, or regulatory compliance.

The development of new or improved waste forms by DOE-EM could offer two potential benefits: (1) more efficient waste processing and immobilization; and (2) enhanced performance of the disposal systems into which the waste forms will eventually be emplaced. With respect to the first benefit, increasing waste loading and/or processing rates could lower production costs and accelerate cleanup schedules (see Chapters 2 and 4). With respect to the second benefit, enhanced performance of the waste form, established under relevant disposal system conditions, could eliminate unwarranted conservatisms and provide greater confidence in the overall performance of the disposal system.

It is important to recognize that repository performance is an optimization problem, and the waste form is one of several elements in the optimization. Other elements include the physical and chemical characteristics of

the rocks hosting the repository as well as the design and characteristics of other engineered barriers. In other words, there is no single figure of merit for waste form performance.

PA can provide safety-based insights to guide future decisions on further development of waste forms. The most notable application relates to estimating thresholds at which extremely low waste-form dissolution rates would control (and simplify the calculation of) the release of all radionuclides for a given disposal environment and design concept. As noted elsewhere in this chapter, the necessary threshold value for waste form durability (expressed as a fractional dissolution rate) to control the performance of a disposal system might be on the order of 10^–6 to as low as 10^–10 per year, depending on factors such as disposal system design, environmental conditions, applicable radioelement solubility limits, and radionuclide loadings in the waste form.

It must be stressed, however, that such exceedingly low dissolution rates for waste forms under disposal conditions are not requirements. Current safety assessments of disposal systems (e.g., Andra, 2005; DOE, 2008; JNC, 2000; Nagra, 2002; SKB, 2006) for a wide variety of HLW and LLW waste forms in various geological formations (e.g., salt, granite, tuff, clay) generally show wide margins of compliance with applicable regulatory safety standards.

As presented in NRC (2003) and illustrated in Figure 7.1, the development of a multiple-barrier repository concept is initially based on (1) identification of waste forms meeting specific waste acceptance criteria (WAC, see Chapter 8); (2) establishment of national laws and regulations (Chapter 8); and (3) selection of a specific disposal site (Chapter 6). Based on this design and collection of field and laboratory data, a series of PAs can be made and the results used to adapt the design of the disposal facility and the iterative collection of additional data, as necessary. From such analyses, the disposal system concept can be adapted and new data collected in iterative stages as necessary. Flexible, staged-adaptation in disposal system development leading to eventual licensing is being implemented in numerous national programs worldwide (NRC, 2003).

This inherent adaptability would also apply to a situation in which a new waste form might be proposed. The waste form would need to pass WAC regarding its physical form, dimensions, and potential impacts on the site environment (e.g., no introduction of major chemical components that might compromise the safety functions of other barriers). As discussed in Chapter 8, there are no specific long-term waste form performance requirements in the WAC; short-term (seven-day) leach rates are used instead as a measure of quality assurance and product consistency. The absence of such performance requirements on waste forms means that an adaptive repository program should readily accommodate new waste forms through

the iterative process of modifying the repository design and updating performance assessment, as illustrated in Figure 7.1. In the case where the calculated releases from a disposal system meet safety criteria because of radioelement solubility limits, then the motivation for developing advanced waste forms would be based more on factors such as waste loading, favorable chemical environment during and after waste form dissolution/alteration that assures such a low solubility, and ease of fabrication, rather than durability.

REFERENCES

Ahn, J., D. Kawasaki, and P. L. Chambré. 2002. “Relationship among Performance of Geologic Repositories, Canister-Array Configuration, and Radionuclide Mass in Waste,” Nucl. Tech. 126, 94-112.

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