Waste Forms Technology and Performance: Interim Report (2010)

The Committee on Waste Forms Technology and Performance (Attachment B) was appointed by the National Research Council in May 2009 to examine requirements for waste form (Box 1) technology and performance in the context of the disposal system in which the waste will be emplaced. The complete statement of task for this study is given in Box 2.

The Department of Energy, Office of Environmental Management (DOE-EM) requested this interim report to provide timely information for fiscal year 2011 technology development planning. The committee has focused this interim report on opportunities associated with selected aspects of the last three bullets of its statement of task (Box 2). These tasks are:

• The state-of-the-art tests and models of waste forms used to predict their performance for time periods appropriate to their disposal system.¹

• Potential modifications of waste form production methods that may lead to more efficient production of waste forms that meet their performance requirements.

• Potential new waste forms that may offer enhanced performance or lead to more efficient production.

The committee judges that the opportunities identified in this report are sufficiently mature to justify consideration by DOE-EM as it plans its fiscal year 2011 technology development program.

In addressing these charges, the committee has focused primarily, but not exclusively, on high-level radioactive waste (HLW) cleanup, which is the longest schedule, highest cost, highest risk, and arguably DOE-EM’s most difficult technical cleanup challenge (see, for example, DOE, 1998, 2010a; NRC, 2001, 2006). At present, tank waste retrieval and closure are limited by schedules for treating and immobilizing HLW in the Defense Waste Processing Facility, which is currently operating at the Savannah River Site; the Waste Treatment Plant, which is under construction at the Hanford Site; and a facility to be designed and constructed at the Idaho Site. Accelerating schedules for treating and immobilizing HLW by introducing new and/or improved waste forms and processing technologies could also accelerate tank waste retrieval and closure schedules.

The committee used its expert judgment to identify the opportunities described in this report. This judgment was informed through a series of briefings, site visits, and a scientific workshop. The committee received briefings on DOE’s current programs and future plans for waste processing, storage, and disposal from DOE-EM, national laboratory, and contractor staff, including information on comparable international programs. The committee visited the Hanford Site (Washington), Idaho Site, Savannah River Site (South Carolina), and their associated national laboratories (Pacific Northwest National Laboratory, Idaho National Laboratory, and Savannah River National Laboratory, respectively) to observe DOE’s waste processing and waste form production programs and to hold technical discussions with site and laboratory staff. The committee also organized a workshop to discuss scientific advances in waste form development and processing. This workshop, which was held in Washington, D.C., on November 4, 2009, featured presentations from researchers in the United States, Russia, Europe, and Australia. The workshop agenda is provided in Attachment C.

A major focus of the DOE-EM cleanup program is on retrieving legacy wastes resulting from nuclear weapons production and testing and processing them into waste forms suitable for disposal in onsite or offsite facilities. Some waste requires minimal processing to make it suitable for disposal; for example, lightly contaminated solid waste generated during facility decommissioning may be suitable for disposal in near-surface engineered facilities with little or no processing. Other waste will require more extensive processing to make it suitable for disposal; for example, HLW, liquid wastes from facility decontamination, contaminated resins from groundwater cleanup, and radioactive sources and other nuclear materials used in civilian and defense applications may require processing to destroy organic components; to remove components that are incompatible with the processing method or final waste form or that are not acceptable for disposal; and to immobilize radioactive and other hazardous components. DOE-EM is using a variety of waste forms to immobilize these components.

The committee observes that the DOE-EM cleanup program is successfully processing waste and producing waste forms at several sites. For example, DOE has completed HLW vitrification at the West Valley, New York, site. DOE is also retrieving HLW from tanks at the Savannah River Site, separating it into high-activity and

low-activity waste streams, and processing these waste streams into high-activity waste glass for disposal in a future geologic repository and low-activity waste saltstone for near-surface onsite disposal. However, DOE-EM’s cleanup program is not expected to be completed for at least another four decades. Consequently, as this program continues, DOE-EM will have opportunities to incorporate emerging developments in science and technology on waste forms and waste form production technologies into its baseline approaches to increase program efficiencies, reduce lifecycle costs and risks, and advance scientific understanding of and stakeholder confidence in waste form behavior in different disposal environments. In short, scientific advances, both now and in the future, will offer the potential for better solutions to DOE-EM’s waste management challenges. It may be important for DOE-EM to maintain sufficient flexibility in its cleanup program to take advantage of these advances.

Based on an analysis of the information it has gathered, the committee observes that waste form science and technology have advanced significantly over the past three decades. The committee judges that there are opportunities to apply these advances in the DOE-EM cleanup program, both now and in the future, to reduce schedules, costs, and risks. The committee offers several observations about potential opportunities in this interim report. Detailed findings and recommendations will be provided in the committee’s final report.

Waste form-relevant science and technology are advancing rapidly along several fronts—for example, chemical and materials processing in industry, waste management in advanced nuclear fuel cycle programs, and management of special nuclear materials in national security applications. There have been numerous recent reports on the development of waste forms and processing technologies for advanced nuclear fuel cycles; some examples are given in Attachment D. Examples of these technologies include:

• Waste form materials designed for significantly higher waste loadings or for improved performance in specific disposal environments.

• Waste processing technologies that can handle large volumes of highly radioactive wastes or that produce highly uniform waste form products.

• Advanced analytical and computational techniques that can be used to understand and quantitatively model interactions between waste forms and near-field² environments of disposal facilities.

Many of these technologies are potentially applicable to DOE-EM waste streams. However, not all are ready for full-scale implementation.

This interim report and the committee’s final report provide only snapshots of these advances. To take full advantage of future scientific and technological

advances, DOE-EM will need to identify, develop where needed, and incorporate where appropriate state-of-the-art science and technology on waste forms and waste form production processes, especially for high-cost, high-risk, and/or orphan³ waste streams. DOE-EM can become cognizant of scientific and technological advances by collaborating with appropriate governmental, scientific, and technical organizations to identify waste forms and waste form production technologies that are potentially applicable to DOE-EM waste streams. For example, collaborations can be established with other DOE offices,⁴ especially the Office of Science and Office of Nuclear Energy; other government agencies (e.g., Department of Defense); scientific, academic, and industrial organizations; and especially other nations’ radioactive waste management programs.

DOE-EM is operating its cleanup program under various regulatory requirements and legal agreements with states and the U.S. Environmental Protection Agency. Modifications of existing requirements or agreements might be necessary before DOE-EM could implement the technologies identified in this report. However, it is outside of the committee’s scope to consider how the use of the technologies identified in this report might impact those requirements and agreements.

The committee has identified four opportunities consistent with its statement of task (Box 2):

• Production of crystalline ceramic⁵ waste forms using fluidized bed steam reforming

• Production of glass, glass composite, and crystalline ceramic waste forms using cold crucible induction melters

• Production of glass, glass composite, and crystalline ceramic waste forms using hot isostatic pressing

• Evaluation of the long-term durability of new waste form materials using experimental studies, laboratory tests, and model development

The first three opportunities involve new applications of existing technologies to DOE-EM waste streams. These waste form production technologies are being used commercially and appear to be applicable for processing and immobilizing a range of DOE-EM waste streams, especially HLW streams. DOE-EM is already planning to apply these technologies to some of its waste streams, as discussed in the following sections. The committee concurs with DOE-EM about the applicability of these technologies and offers observations in this interim report on the wider application of these technologies in the cleanup program.

The fourth opportunity involves extending the application of experiments, tests, and model development for evaluating the durability⁶ of new waste form materials over time periods for concern for disposal (typically 10³-10⁶ years). This would provide DOE-EM with future flexibility to use new waste forms in its cleanup program and enhance the long-term safety of disposal.

Fluidized Bed Steam Reforming (FBSR; see Attachment E for a brief technology description) is a robust technology for processing wastes. Its primary advantages are high throughput and ability to accommodate a wide range of feeds and additives, including feeds containing anionic sulfur and nitrogen species, halides, and organics that are incompatible with some other types of waste forms and waste form production processes.

FBSR is based on fluidized bed technology, which was invented in the 19^th century and found widespread use in the refining and chemical industries starting around World War II. Applications of fluidized bed technology in nuclear fuel production, fuel recovery, and waste processing date back to late 1950s and early 1960s. For example, fluidization was used for the reduction and hydrofluorination of uranium concentrates and calcination of high-level radioactive waste. Two calcination facilities were successfully operated at the Idaho National Engineering Laboratory (now Idaho National Laboratory) from 1963 to 1981 and from 1981 to 2000 to immobilize HLW.

The FBSR process is already being used commercially for processing nuclear waste. A commercial facility to continuously process organic radioactive wastes at moderate temperatures in a hydrothermal steam environment was built by Studsvik in Erwin, Tennessee, in 1999. The Erwin facility uses a steam reforming technology, referred to as THermal Organic Reduction (THOR^®), to pyrolyze organic resins loaded with Cs-137 and Co-60 from commercial nuclear facilities. The Erwin facility has the current capability to process a wide variety of solid and liquid streams including ion exchange resins, charcoal, graphite, sludge, oils, solvents, and cleaning solutions at radiation levels of up to 400R/hr (Mason et al., 1999).

FBSR is not a combustion process and is Clean Air Act (CAA) compliant. It has also been shown to be Hazardous Waste Combustor (HWC) Maximum Achievable Control Technology (MACT) compliant for mercury, chlorine, carbon monoxide, total hydrocarbons, and heavy metals. A significant benefit of the FBSR process is that liquid secondary wastes are not produced (Mason et al., 1999). Many years of operating and design experience with fluidized beds in the chemical industry and the availability of computational fluid dynamics tools significantly reduce development and operating risks for potential EM applications.

Depending on the starting material feeds, FBSR produces a range of waste form compositions. If kaolinite is added to an alkali-rich waste (e.g., the low-activity waste fraction of neutralized HLW) during processing,⁷ a crystalline ceramic waste form is produced that is composed of Na-Al-Si feldspathoid mineral analogs (e.g., sodalite) that serve as potential hosts for a number of radionuclides (Attachment E). Bench scale, pilot scale, and engineering-scale tests have all produced this mineral assemblage using a variety of DOE waste simulants as feed materials. Additionally, an illite-type clay additive has been tested at the bench scale and shown to form dehydroxylated mica, which is a good host for lanthanides, cesium, strontium, barium, rubidium, and thallium (Jantzen and Williams, 2008). It is reasonable to expect that these mineral assemblages would also serve as hosts for the radioactive forms of these elements that are present in DOE-EM waste streams.

DOE-EM plans to apply FBSR to some of its waste streams. An FBSR facility is being designed and constructed at the Idaho Site for treatment of decontamination solutions (referred to as sodium-bearing waste) for potential disposal in the Waste Isolation Pilot Plant (Marshall et al., 2003). Another facility is being designed for use at the Savannah River Site to process HLW in Tank 48, which contains nitrates, nitrites, and organic sodium tetraphenyl borate (NaTPB). This process will produce carbonate or silicate phases which can be fed to the Defense Waste Processing Facility for vitrification (Jantzen, 2004). DOE-EM has also carried out pilot-scale testing on a variety of simulated wastes to produce aluminosilicate ceramic waste forms.

The committee observes that there are at least two potential types of applications of FBSR in the DOE-EM cleanup program:

1. As a front-end process for conditioning waste feed streams:

   • For accelerating liquid evaporation at the front end of the HLW vitrification process, which could enable increased waste throughputs to the Joule-heated melters (see Attachment F) and increased production rates of high-activity and low-activity waste forms.

   • For processing waste streams, including resins, containing large quantities of organic materials and nitrates. The planned application of FBSR to process Savannah River tank waste containing high concentrations of NaTPB is an example of such an application. FBSR also has potential applications for processing waste streams containing

• organic solvents and radionuclide-loaded organic resins, for example, the technetium-99-loaded resins generated by groundwater cleanup efforts at the Hanford Site.

2. As a process for production of crystalline ceramic waste forms:

   • For processing alkaline HLW with bulk aluminosilicate additives (e.g., kaolinite), which could produce waste forms with good radionuclide retention properties and waste loadings comparable to or greater than borosilicate glass (Jantzen, 2006). This process could also reduce or eliminate the need for effluent recycling. This process is potentially applicable to both high-activity and low-activity waste streams and in fact has been demonstrated at the pilot scale on Hanford waste simulants (see www.thortt.com for technical papers).

   • For processing recycle liquids from HLW waste processing operations. This application has already been demonstrated at pilot scale for low-activity secondary waste simulants at Hanford.

FBSR is a mature technology. Its deployment in DOE-EM applications may require some up-front development work to tailor it to specific waste streams, but relatively little basic research is likely to be required. For example, development work might be required to better understand and ameliorate the attrition of granular bed material present in FBSR. Such attrition can be reduced through development work that is focused on the proper design of internal components, dust collection equipment, operating conditions, and selection of additive materials. All of these have well known solutions in the chemical or petroleum industry applications of fluidized beds.

Any waste forms produced using FBSR must, like all other waste forms, undergo characterization work to understand key structural characteristics, for example, how radionuclides are incorporated into atomic structures. See the section entitled “Waste Form Durability: Experiments, Tests, and Model Development” for additional discussion of this issue.

The Cold Crucible Induction Melter (CCIM) is a promising technology for producing glass, glass composite, and crystalline ceramic waste forms. CCIMs are potential replacements for Joule-heated melters (JHMs; see Attachment F), which are part of the current DOE baseline for production of high-activity and low-activity waste glass.

A CCIM consists of water-cooled metal tubes that are arranged to form a crucible. An inductor surrounding the crucible produces a high-frequency alternating current that induces eddy currents (and resultant Joule heating) of materials contained in the crucible. The melting process is usually initiated by inserting a resistive heating element into the crucible to obtain sufficient melt to couple with the electromagnetic field. At that point, the resistive element can be removed so that no foreign materials are in contact with the melt. A solid “skull” of quenched waste material, typically a few

millimeters in thickness, forms along the crucible wall, protecting it from degradation and corrosion.

CCIMs have several advantages over both current-generation and advanced JHMs. They allow for higher throughputs and waste loadings. They are operationally simpler and allow for faster recoveries from system upsets.⁸ The absence of internal electrodes and refractories allows for increased melter longevity and permits higher-temperature operation compared with current-generation JHMs. As a consequence, CCIMs can be used to process a wider range of waste compositions, including corrosive wastes that are incompatible with current-generation JHMs. Additionally, they can more easily accommodate differing glass compositions, including iron phosphate glasses, that are incompatible with some JHM internal components. CCIMs can be cycled frequently with varying feed compositions without thermal damage or loss of compositional control. And they are capable of producing crystalline ceramics through controlled crystallization.

CCIM is a flexible processing method that can be used in conjunction with other technologies. For example, an integrated process that combines an oxygen plasma and induction-heated cold crucible is reported by Vernaz and Poinssot (2008). This process, which is still under development, is referred to as the Advanced Hybrid System for Incineration and Vitrification (SHIVA). It consists of a single reaction vessel which has three functions: (1) incineration (2) vitrification, and (3) gas post-combustion. This is a promising technology for processing wastes containing radioactive, organic, and other hazardous chemical constituents that are difficult to separate by other processes. The plasma decomposes organic material, significantly reducing its volume, and produces a high-quality containment material (glass in this instance).

CCIM development began in France and Russia in the 1970s (Elliott, 1996). The Russians are using CCIMs to process radioactive waste at the Mayak Plant, and the French are using a CCIM to vitrify HLW at an industrial scale at the La Hague plant. DOE-EM is currently investigating CCIM technology for possible use in its HLW immobilization programs.

CCIM is a mature technology for the vitrification of fission product solutions and decontamination waste streams. It also has potential applications for processing metallic waste streams (Vernaz, 2009). The underlying technology is proven, but operational experience in large-scale waste stream processing environments is limited in comparison to JHMs. Its deployment in DOE-EM applications may require some up-front development work to ensure its compatibility with specific process flowsheets, but no basic research is likely to be required.

Because CCIMs are smaller per unit of throughput and operationally more robust than JHMs, they could potentially be back-fitted to the Defense Waste Processing Facility at Savannah River and the Waste Treatment Plant at Hanford. For example, IPET (2003) examined the feasibility of replacing the JHMs in the Waste Treatment Plant at Hanford with CCIMs. It concluded that two CCIMs could be retrofitted into each of the

two melter cells in the plant. If the melters were installed before the plant was hot commissioned, about 4 months would be required to modify the melter cells and install the new equipment (IPET, 2003, p. 4.70). Additional time would be required to install the melters after hot commissioning—either to decontaminate the melter cells prior to installation of the new equipment or to construct a new melter facility.

Hot isostatic pressing (HIP) produces waste forms through the application of heat and pressure. The waste and other materials to be processed are loaded into a can, which is welded shut and placed in a pressure vessel inside an electrically heated furnace. The loaded can is heated and a high isostatic pressure is applied, which compresses the waste into a solid, monolithic waste form.

The HIP process, originally referred to as gas-pressure bonding, was first developed by Battelle Memorial Institute in the mid 1950’s (ASME, 1985). Its initial use was for manufacturing nuclear fuels, but it is now a well-established technology used by a wide range of industries for castings, tool making, and manufacturing of ceramic components. The Australian Nuclear Science and Technology Organisation (ANSTO) has developed and demonstrated HIP for immobilizing radioactive wastes from medical isotope production; it plans to commence the commissioning of a HIP facility for this purpose at the end of 2011 (Kath Smith and Bruce Begg, ANSTO, written communication). HIP has never been used to immobilize nuclear waste in the United States. In January 2010, DOE announced its formal decision to use HIP to convert the HLW calcine at the Idaho Site into ceramic-like waste forms (DOE, 2010b). However, the technology readiness assessment is still in progress and a safety assessment has not yet been completed.

HIP is a mature and safe technology as demonstrated by its wide use outside the nuclear industry. The pressure vessels are designed with stringent codes such as those developed by the TÜV (Technischer Überwachungs-Verein [Technical Inspection Association], a German product safety and quality assurance testing firm) and the American Society of Mechanical Engineers (ASME). The conservative ASME code and inspection regime are designed to ensure that vessel integrity is maintained over its service life. Other safety features include active and passive over-pressure control systems and safety shields.

HIP also has many potential advantages for processing nuclear waste. Notably, it produces monolithic waste forms with substantially reduced volumes compared to untreated waste streams. Because the waste is processed in a sealed can, there are no volatile emissions. Also, there is no direct contact between the waste and the HIP apparatus, so secondary waste generation is minimized. HIP is compatible with a wide range of waste compositions, although it has a limited tolerance for gases and volatiles. It can produce glass, glass composite, and crystalline ceramic waste forms.

Unlike many other consolidation technologies, HIP does not require stringent control of physical properties such as viscosity, melt temperature, or melt conductivity, therefore permitting significantly higher waste loadings. However, it does require that the waste form additives be tailored to sequester radionuclides in specified host phases.

For production of SYNROC,⁹ for example, REDuction/OXidation (REDOX) conditions must be controlled to form the desired phase assemblages. In addition, processing conditions (pressure and temperature) must be closely controlled.

Although HIP is a flexible technology it does have some limitations. Crystalline ceramic waste forms produced by HIP (as well as conventional press and sinter technology) may contain intergranular glassy phases, especially when incorporating waste containing alkali or alkaline earth species in the presence of glass formers such as silicon or boron. This intergranular glass may limit product durability (e.g., Clarke, 1981; Cooper et al., 1986; Zhang et al., 2010). Also, HIP has been demonstrated only at small scales to date. The small size of the waste cans and long times required for heating currently limits the application of this technology to volumetrically small waste streams.

Given its flexibility, HIP is potentially applicable to a range of DOE-EM waste streams, including orphan waste streams (see Footnote 3) whose diversity requires versatile methods for treatment and immobilization, as well as waste streams that are difficult or inefficient to process by other technologies because of physical or chemical heterogeneity. However, additional studies are needed to demonstrate the safety and compatibility of this technology with specific waste streams and to address its scalability to high-volume waste streams.

As discussed in Box 1, the primary function of a waste form is to sequester radioactive or hazardous constituents in stable, solid matrices either by chemical incorporation or encapsulation. Demonstrating that a given waste form has sufficient durability (see Footnote 6 for a definition) to perform this function over the long time periods of concern for disposal (typically 10³-10⁶ years as noted previously) is a scientific and technical challenge and arguably presents a major obstacle to stakeholder acceptance of waste form disposal strategies. The primary challenge involves extrapolating the durability behavior observed in short-term laboratory tests to these longer time scales—that is, evaluating long-term durability of waste forms.

Short-term (typically days to months) laboratory tests¹⁰ cannot be used directly to evaluate long-term durability. Such evaluation requires the establishment of parallel but connected programs of experimental studies, laboratory tests, and model development tailored to specific combinations of waste forms and disposal environments:¹¹

• Experimental studies¹² are used to identify the mechanism(s) of waste form alteration and release of contained radioactive and other hazardous constituents of concern. This information is used to develop mechanistic models that account for the important physical and chemical processes that govern waste form alteration and radionuclide release. Studies of natural analogs—e.g., glasses and ceramics that have survived in natural environments for thousands of years—can also provide useful information on release mechanisms that operate over long time periods and might not be observed in the laboratory.

• Laboratory testing is used to measure short-term release rates of radioactive and hazardous constituents from the waste form and the formation of reaction products. The American Society of Testing and Materials (ASTM) has developed a suite of tests that can be used to measure release rates for some types of waste form materials, including certain glass, glass composite, glass ceramic, and metallic materials. However, these tests are not suitable for application to all waste form materials.

• Coupled models are used to evaluate long-term waste form durability for a given set of near-field conditions in a disposal environment. These models couple the mechanistic release models described above with transport models that account for the movement of radioactive and hazardous constituents (which can be in dissolved, colloidal, or gaseous form) from the altered waste form into the near-field environment of the disposal facility (Steefel et al., 2005). The parameters used in these coupled models are frequently derived from laboratory tests.

Key features of the coupled models can also be abstracted to develop performance models of the disposal system. These models are used to assess the long-term performance of the disposal system, which is usually expressed in terms of an annual dose to maximally exposed individuals who live near the facility at some specified future time.¹³ The waste form and other engineered and natural barriers in the disposal system are intended to function together to reduce these doses by retarding the release of radionuclides and other hazardous constituents from the facility.

As the foregoing discussion suggests, experimental studies, laboratory testing, and modeling work proceed hand-in-hand: the experimental results are used to identify appropriate tests; the experiments and tests inform the modeling work; and the modeling work uncovers additional needs for information that inform further experimental and testing work. This development work usually requires considerable investment of time, especially if testing protocols need to be developed, modified, or qualified for new materials.

The long-term durability of borosilicate glass, the waste form material being used by DOE-EM to immobilize HLW, has been established using the approaches described above. DOE-EM has also used borosilicate glass as a reference standard for benchmarking the durability of other waste form materials. For example, at the Hanford Site, DOE-EM and the state of Washington have agreed that the waste form selected for immobilizing low-activity waste for near-surface disposal must be “as good as glass,” meaning that it must have at least the same durability as the Low Activity Waste Reference Material (LRM), a borosilicate glass developed for Hanford’s low-activity waste.

On the other hand, DOE-EM has identified orphan waste streams that have no identified disposition pathways (see Footnote 3). Once waste forms for such waste streams are identified, their long-term durability would need to be established using the methods that are described in this section.

Demonstrating the durability equivalence of different waste form materials is not a simple matter. The tests used to evaluate the durability of one waste form material cannot be applied to another waste form material unless it can be demonstrated that both materials undergo alteration and radionuclide release by the same mechanism(s). For example, tests used for evaluating the long-term durability of borosilicate glass are applicable, either directly or with some modification, to some other types of glass, glass composite, and glass-ceramic waste forms. However, by themselves, they are probably not applicable—or are inadequate—to establish the durability equivalence of other waste form materials such as cements, hydroceramics, or geopolymers. Moreover, some test measurements that are critical for estimating release rates, for example surface area measurements, have not been standardized for waste form materials such as foam glass¹⁴ and crystalline ceramics. The final judgment of the suitability of a waste form for any specific geologic environment will inevitably be the result of a combination of the results of standard tests, experiments, models, as well as confirmatory field and analog studies, as outlined, for example, in the ASTM Standard C 1174-07 (see Footnote 11).

The durability comparison also requires the specification of the near-field conditions (e.g., water flow rate, porewater composition, partial pressure of CO₂, and pH) in the disposal facilities that will host the waste forms.¹⁵ The same waste form might exhibit widely different durability behaviors if they are disposed of in facilities having different near-field conditions. For example, durability of a waste form in wet environments can be tested as a function of pH using buffer solutions as leachants.

The assessment of long-term durability will be required for any new waste forms that DOE-EM intends to use in its cleanup program. In the committee’s judgment, assessment of the long-term durability of crystalline ceramic waste forms represents a key near-term opportunity for DOE-EM. Crystalline ceramic materials produced, for example, by FBSR and HIP have been identified elsewhere in this interim report as flexible waste forms with many potential applications, including high-activity and low-activity waste immobilization. Evaluating the long-term durability of these materials for a

variety of near-field conditions could provide future flexibility to apply them more widely throughout the clean-up program.

As noted previously, this interim report has focused on near-term opportunities that the committee judges will be useful to DOE-EM for planning its fiscal year 2011 technology development programs. The committee’s final report will address the statement of task in its entirety. It will provide a more detailed assessment of waste forms, processing technologies, and state-of-the-art tests and models. It will also identify longer-term research, development, and deployment opportunities for DOE-EM’s cleanup program. The final report is scheduled for completion in September 2010.