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Introduction

Vitrification of liquid high-level radioactive waste (HLW) has received greater attention, worldwide, than any other high-level waste solidification process. The industrial-scale demonstration of vitrification dates from the operation of the AVM (Atelier de Vitrification de Marcoule ) plant in Marcoule, France, in 1978. In the United States, the Defense Waste Processing Facility (DWPF) at the Savannah River Site in South Carolina and the West Valley Demonstration Project (WVDP) in New York both began operation with radioactive waste in 1996. The DWPF is designed to solidify defense HLW (over 130,000 m3 of HLW from 51 tanks), and the WVDP is designed to solidify high-level waste from commercial reprocessing of spent nuclear fuel (2,200 m3) (see abstracts by S. P. Cowan and by W. Lutze in Appendix E).

Borosilicate glass is a preferred waste form in most countries. In the former Soviet Union, a phosphate glass has been developed for its two vitrification plants, but, as reported in May 13-15, 1996, at the National Research Council International Workshop on Glass as a Waste Form and Vitrification Technology, Russia is developing vitrification technologies that also will use a borosilicate glass, as well as other alternative waste forms for separated long-lived radionuclides (see abstract by Aloy et al. in Appendix E).

With a long history of industrial-scale success, particularly in France, where over 6,000 canisters have been filled with vitrified high-level waste at La Hague, and with the general acceptance of borosilicate glass as a waste form, one might ask what the purpose of convening the workshop was? The simple answer is that, despite the present success of vitrification, much lies ahead in the field of radioactive waste management. In the United States vitrification is proposed or planned at numerous sites, such as the Fernald Site in Ohio and the Hanford Site in Washington. Vitrification is proposed not only for HLW, but also for the very large volumes of low-level waste found throughout the U.S. Department of Energy (DOE) defense complex. Most recently, a report of the National Academy of Sciences (1994) described vitrification as one of the "promising" options for the disposition of excess weapons plutonium (up to 50 metric tons in the United States).

The scale oft he vitrification projects is immense. At the Hanford site alone, the inventory includes 11 million m3 of fluids (of this, 216,000 m3 are high-level waste fluids and sludge) and 6,900 metric tons of nuclear materials, which include 4,100 metric tons of uranium and 15 metric tons of cesium and strontium capsules. The complexity of the tank waste defies simple compositional classification (National Research Council, 1996). With approximately 40 chemical classifications for the high-level waste tanks at Hanford, over one-third of the tanks are chemically unique in that they do not fit any classification (Bunker et al., 1995; see also Appendix D).

Furthermore, the investment in time and dollars can be substantial. For instance, the DWPF was originally scheduled to begin operation in 1989 at a projected construction and startup cost of approximately $1 billion. Instead, the first canisters were poured in 1996, and projections by the General Accounting Office (1992) suggest that the cost may reach $4 billion. The size of the investment and the time required to bring a vitrification facility into operation require a careful review of past practice and present knowledge in order to take advantage of the full range of future possibilities.



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Glass as a Waste Form and Vitrification Technology: Summary of an International Workshop 1 Introduction Vitrification of liquid high-level radioactive waste (HLW) has received greater attention, worldwide, than any other high-level waste solidification process. The industrial-scale demonstration of vitrification dates from the operation of the AVM (Atelier de Vitrification de Marcoule ) plant in Marcoule, France, in 1978. In the United States, the Defense Waste Processing Facility (DWPF) at the Savannah River Site in South Carolina and the West Valley Demonstration Project (WVDP) in New York both began operation with radioactive waste in 1996. The DWPF is designed to solidify defense HLW (over 130,000 m3 of HLW from 51 tanks), and the WVDP is designed to solidify high-level waste from commercial reprocessing of spent nuclear fuel (2,200 m3) (see abstracts by S. P. Cowan and by W. Lutze in Appendix E). Borosilicate glass is a preferred waste form in most countries. In the former Soviet Union, a phosphate glass has been developed for its two vitrification plants, but, as reported in May 13-15, 1996, at the National Research Council International Workshop on Glass as a Waste Form and Vitrification Technology, Russia is developing vitrification technologies that also will use a borosilicate glass, as well as other alternative waste forms for separated long-lived radionuclides (see abstract by Aloy et al. in Appendix E). With a long history of industrial-scale success, particularly in France, where over 6,000 canisters have been filled with vitrified high-level waste at La Hague, and with the general acceptance of borosilicate glass as a waste form, one might ask what the purpose of convening the workshop was? The simple answer is that, despite the present success of vitrification, much lies ahead in the field of radioactive waste management. In the United States vitrification is proposed or planned at numerous sites, such as the Fernald Site in Ohio and the Hanford Site in Washington. Vitrification is proposed not only for HLW, but also for the very large volumes of low-level waste found throughout the U.S. Department of Energy (DOE) defense complex. Most recently, a report of the National Academy of Sciences (1994) described vitrification as one of the "promising" options for the disposition of excess weapons plutonium (up to 50 metric tons in the United States). The scale oft he vitrification projects is immense. At the Hanford site alone, the inventory includes 11 million m3 of fluids (of this, 216,000 m3 are high-level waste fluids and sludge) and 6,900 metric tons of nuclear materials, which include 4,100 metric tons of uranium and 15 metric tons of cesium and strontium capsules. The complexity of the tank waste defies simple compositional classification (National Research Council, 1996). With approximately 40 chemical classifications for the high-level waste tanks at Hanford, over one-third of the tanks are chemically unique in that they do not fit any classification (Bunker et al., 1995; see also Appendix D). Furthermore, the investment in time and dollars can be substantial. For instance, the DWPF was originally scheduled to begin operation in 1989 at a projected construction and startup cost of approximately $1 billion. Instead, the first canisters were poured in 1996, and projections by the General Accounting Office (1992) suggest that the cost may reach $4 billion. The size of the investment and the time required to bring a vitrification facility into operation require a careful review of past practice and present knowledge in order to take advantage of the full range of future possibilities.

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Glass as a Waste Form and Vitrification Technology: Summary of an International Workshop Finally, despite many years of development and large investments in nuclear waste solidification, the present technologies are essentially of only two types; a single-stage liquid-fed ceramic melter and a two-stage rotating calciner and a metallic melter, both producing a borosilicate glass. Are there new technologies? Are there other waste forms? Furthermore, vitrification plants are expected to be operational for the next 30 to 50 years. As in any large-scale industrial endeavor, both the process and the product will certainly be improved with time. Any plant modifications must be based both on practical operational experience and a sound scientific understanding of fundamental issues and phenomena. The National Research Council (NRC) workshop was aimed at identifying and discussing these underlying fundamental issues. Given the scale and complexity of the task of vitrifying nuclear wastes, it is prudent to review the past 30 years of successes and failures and to use this knowledge to plan an efficient path for the very large job ahead. Future plans should reflect the best application of processing technologies and a knowledge of waste forms that reduces risk to the public and the environment while also avoiding unnecessary cost and delay. ORGANIZATION OF THE WORKSHOP In light of the above, the NRC Board on Radioactive Waste Management decided that it was timely to convene an International Workshop on Glass as a Waste Form and Vitrification Technology. The DOE Waste Management Program provided financial support and logistical assistance for the workshop. A general goal of the workshop was to review the current state of knowledge of glass as a waste form for the immobilization of radioactive wastes. The workshop was held on May 13-15, 1996, in the auditorium of the National Academy of Sciences in Washington, D.C., and was attended by approximately 250 participants (see Appendix C) representing 12 countries. The workshop program, given in Appendix B, included sessions on the Present Status of Vitrification, Glass Durability and Modeling, and Operational Experiences with Vitrification, as well as posters, panel discussions, and summaries by rapporteurs. This report summarizes some of the major discussions of the workshop from the perspective of the steering committee; extended abstracts of the presentations are included in Appendix E. A three-day program cannot capture the detail of highly specialized waste management meetings held around the world (of which there are many; specifically the reader is referred to the proceedings of the Materials Research Society's annual symposium on the Scientific Basis for Nuclear Waste Management, the biennial symposium of the American Society of Mechanical Engineers, and the annual Waste Management meetings held in Tucson, Arizona). The NRC workshop was not intended to be an exhaustive review of the status of this complex subject, but rather an opportunity to identify remaining critical areas for future research based on presentations by leading international experts in the field. The workshop program (see Appendix B) was constructed around the four following issues that were addressed by the invited speakers. Waste Characteristics: What are the volumes, states, and compositions of the waste that might be vitrified? Regulations: In the United States, what regulations apply to the waste form, and what is the basis for the regulation?

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Glass as a Waste Form and Vitrification Technology: Summary of an International Workshop Waste Form Properties: What is the state of knowledge of the fundamental properties (e.g., chemical and mechanical durability) of the waste form? Technologies: What is the state of practice of vitrification technologies, and what has been the international experience? Since the major purpose of the NRC workshop was to determine whether adequate interactions exist between these four issues, the workshop steering committee scheduled considerable time for discussion after presentations and during the program, and the audience obliged with knowledgeable enthusiasm. There was a clear difference in emphasis and perspective between those responsible for developing and applying the vitrification technologies and those required to evaluate waste form performance in a repository. The technology is driven by the chemical complexity and large volumes of the waste. These chemically heterogeneous radioactive waste streams of large volume must be handled remotely and safely. Research on waste form properties is driven by special attention to long-term behavior (e.g., chemical durability and radiation effects). The regulations view the waste form in a variety of roles: (1) as a medium for safe transport, (2) as a medium for interim storage, or (3) as one of the engineered barrier systems in the geologic repository. The workshop explored the connection between technology, science, and regulation. The adopted technologies should be able to incorporate large and diverse waste streams by remote operation into a final product that satisfies regulatory requirements (e.g., product consistency) and for which there is a scientific basis for the evaluation of its long-term behavior, particularly the waste form's chemical durability. HISTORY As early as the 1950s, glass was considered an important potential waste form for radioactive materials. The technology of glass formation has a long history that can be traced back to ancient times, and borosilicate glasses have been used since early in this century. A historical summary of the development of glass as a waste form and the parallel vitrification technologies can be found in Lutze (1988) and Bates et al. (1994). During the past 20 years, there have been numerous reviews and comparisons of waste forms, which are summarized, in part, by Bates et al. (1994) and in detail by Lutze and Ewing (1988). In general, the positive evaluations of glass as a radioactive waste form and vitrification technologies have rested on the following: As a nonstoichiometric solid, glass can accept a wide range of waste stream compositions. As an aperiodic solid, the structure of glasses is considered less susceptible to radiation damage effects than crystalline materials. As a waste form, in combination with other barriers to radionuclide migration (e.g., the canister, backfill, geology), glass usually is considered to be a more than adequate barrier to radionuclide release. Industrial-scale production of glass incorporating radioactive waste has been demonstrated, and this production experience has been gained in a number of countries.

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Glass as a Waste Form and Vitrification Technology: Summary of an International Workshop Concerns in general about glass as a waste form are based on the fact that "glass corrodes slowly in water and humid air, and inevitably, certain quantities of radionuclides are mobilized. The glass is not inherently corrosion-resistant, but rather depends on the waste package and on surrounding geochemical and hydrological constraints" (Grambow, 1995). The previous reviews of glass and vitrification technologies have at every step failed to present a balance of the experience with industrial-scale vitrification technologies against the demonstrated adequacy of glass as a material for the immobilization of radioactive waste over long periods of time. The NRC workshop brought these two perspectives to the same forum. For most countries the reason for developing vitrification for radioactive waste remediation was permanent disposal of the products of reprocessing. This need led to the development of the two basic approaches, exemplified by the French process (also used in the United Kingdom) and the German process (used in Belgium and planned for China). In 1977 the United States decided not to reprocess commercial spent nuclear fuel, but instead to dispose of it directly from reactors (Carter, 1987). The DOE program focused on vitrifying defense waste at the Savannah River and Hanford sites, as well as waste from reprocessing of commercial fuel at the West Valley Demonstration Project. In the past 10 years, substantial scientific progress has been made in understanding many aspects of glass behavior. During the same period, commercial plants in France and England have been operated, as has a Belgian vitrification plant at Mol. The DWPF at the Savannah River Site began operation in April 1996, and the West Valley Demonstration Project facility began operation in June 1996, after the NRC workshop. Thus, enough is known about processing radioactive waste into glass to build and operate several large-scale facilities. However, the construction and operation of these plants are expensive, and each plant generally has been designed to accommodate a particular range of waste compositions (also known as feedstocks). None have been built to handle the wide range of potential feedstocks represented by the U.S. and Russian waste that have resulted from years of nuclear weapons production. THE U.S. PROGRAM The goal of the U.S. program is to reduce risks to the public and the environment at an acceptable cost. This goal can be approached by getting better performance and, therefore, more risk reduction at today's unit costs (assuming today's cost are acceptable) or accepting today's performance and reducing unit costs (assuming today's performance is acceptable). Ideally, one strives for both improved performance and reduced costs. Presently the United States does not have a repository ready to accept HLW. The first high-level waste that will go into the first U.S. repository, perhaps at Yucca Mountain, Nevada, most likely will be commercial spent fuel, in an as-yet-undecided container, and the glass "logs" being produced at the DWPF in Savannah River and the WVDP in West Valley facilities. Thus, repository designs here and abroad will be tailored to existing glasses and in the United States to the disposal of spent nuclear fuel. More than 90 percent of the curie content of waste scheduled for a U.S. repository is in spent reactor fuel. The shift in U.S. policy in the 1970s from reprocessing to direct disposal was equivalent to determining that oxide fuel is acceptable as a waste form. However, performance criteria for other waste forms have not been determined (e.g., glass "logs" from the DWPF and the West Valley Demonstration Project, or for the final selected waste form for Hanford wastes, which

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Glass as a Waste Form and Vitrification Technology: Summary of an International Workshop constitute the largest volume of HLW in the DOE complex). Neither the U.S. Environmental Protection Agency (EPA) nor the U.S. Nuclear Regulatory Commission (USNRC) currently has published regulations for Yucca Mountain on which acceptance criteria can be based, although EPA is drafting regulations after considering the National Research Council (1995) report Technical Bases for Yucca Mountain Standards . The USNRC is concerned that the performance assessment necessary for licensing will be extremely difficult because of uncertainties in projecting glass performance in repositories over geological time scales. Unvitrified low-level waste must meet EPA and USNRC criteria found in the Code of Federal Regulations Title 10, Part 61 (10 CFR 61), which are based on lined trenches. In principle, meeting these requirements for vitrified waste will not be difficult. The growing interest in in situ vitrification will be more complicated. Because there will be no pretreatment of waste subjected to in situ vitrification, such waste will be considered as mixed waste, and Resource Conservation and Recovery Act (RCRA) regulations will apply. However, large volumes of liquid wastes at Hanford (larger in volume and almost equivalent in total activity to the waste at Savannah River) remain to be treated. How should this waste be vitrified? By what technologies? At what costs? These are important issues. Additionally, improving the understanding of the glass properties that may have an impact on performance remains important. Glass can be viewed from a performance perspective as filling three roles or "futures": (1) as the sole barrier between the radioactive wastes and the biosphere. (2) as a major barrier in a system of multibarriers (e.g., glass, canister, backfill, geology), or (3) as a convenient container for transport and temporary surface storage. The actual role that glass will play depends very much on the behavior of the repository system. Each of these roles is discussed in the next section. Depending on the strategy adopted, different priorities will be assigned to addressing the scientific and technical issues. In all cases, however, the current best knowledge of glass performance can significantly improve assessments of repository performance.