High-level Waste Vitrification Research Articles

Nuclear Waste: the challenge is global

The difficulties in disposing of many millions of tons of nuclear waste, comprising radioactive elements with half-lives ranging from microseconds to eons, are discussed, stressing the global nature of the issue. The IAEA classification of nuclear waste into high, intermediate, and low-level forms is described. The classification is based on several factors, including the waste's source and temperature, and half-life: the greater the number of long-lived radioisotopes the waste contains, the higher the level of waste. Although, overall, wastes from commercial power reactors outweigh those from military ones, military programs produce a huge excess of the high-level type. Methods of reprocessing high-level wastes, the solution most favored in Europe, are examined. Opposition to permanent underground disposal sites, the principal alternative to reprocessing, is discussed. Progress in disposing of low-level wastes is described. The technology of vitrification of high-level wastes and of nuclear fuel reprocessing is examined. >

1989 Joint International Waste Management Conference

Reviewed is the as-titled conference, held in October 23 to 28, 1989, in Kyoto, Japan, sponsored by American Society of Mechanical Engineers, Japan Society of Mechanical Engineers, and Atomic Energy Society of Japan, with other co-sponsoring societies. In the Conference, all the aspects of radioactive waste management were discussed, including policies and general strategies for waste management, solidification and incineration of low-, and intermediate-level wastes (LLW and ILW), shallow-land burial disposal of LLW/ILW and its safety assessment, decommissioning, storage and transportation of spent fuels, vitrification of high-level wastes (HLW), deep geologic disposal of HLW and its safety assessment by natural analogs, and partitioning and transmutation. The significance of the Conference is found in that the informations of Japan's waste management activities, which do not always reach audience in other part of the world, were conveyed in detail in the Conference. More papers were presented for LLW/ILW...

Effects of Composition Variations on Microstructure and Chemical Durability of West Valley Reference Glass.

ABSTRACTGlasses suitable for high-level nuclear waste vitrification must meet a number of requirements including processability, durability, and phase stability. Our new data indicate that the West Valley Reference 5 composition meets these requirements: it is phase stable over the expected range of melter temperatures and residence times, and only minimal (<2 vol %) secondary phase formation, consisting predominantly of iron-group spinels, is expected in the canister cooled glass. Leach tests have shown that the durability of Reference 5 is significantly less sensitive to spinel formation than the earlier Reference 4. However, natural process variations during production will result in a range of glass compositions around the nominal composition. In this paper we report on the effects of composition changes around Reference 5 on the phase stability upon heat treatment and the consequent effects on chemical durability. Since variations in the waste-stream levels of most major components can be accommodated by adjusting the quantities of the glass-forming additives we focus here on a group of components for which this is generally not the case: Ca, Ce, Cr, Mn, Ni, S, and Fe. While many of these components are present at below 1 wt % they could have potentially significant indirect effects on product durability due to enhancement of secondary phase formation. The results provide a data base to support broad (≥300%) tolerance ranges to variation of levels of most of these components. Viscosity data are also reported.

Product Models for the Vitrification of West Valley High-Level Wastes

ABSTRACTProperty models have been developed for the major properties that need to be controlled in the production of borosilicate glasses for West Valley high-level nuclear waste immobilization. The chemical durability is the most important parameter for product performance, while melt viscosity is the most critical parameter in assuring the processability of the glass. Simple models for these properties are described that are based on data from numerous glasses which were prepared with compositions in the region around the West Valley reference glass. A scheme for optimization of the target glass and for predicting the acceptability of glasses resulting from natural process variations is illustrated. This involves integration of the product models with a process model that was described previously. This approach has guided the present placement of the West Valley reference glass.

Influence of iron ions on the structural properties of some inorganic glasses

The effects of iron on the structural properties of Zn-borosilicate glass and Pb-metaphosphate glass were studied using X-ray diffraction,57Fe Mossbauer spectroscopy and IR spectroscopy. Zn-borosilicate glass was prepared with varying amounts of Fe2O3 (up to 30% wt.). It was found that the chemical form of added iron (α-FeOOH, α-Fe2O3 or Fe3O4) affects the Fe3+/Fe2+ ratio, as well as the distribution of iron ions at different coordination sites. At high concentration of iron the crystallization of zinc ferrite in the glass matrix takes place. X-ray diffraction and57Fe Mossbauer spectroscopy showed that the amount of zinc ferrite in Zn-borosilicate glass decreases with the following order of addition: α-FeOOH→α-Fe2O3→Fe3O4. In Pb-metaphosphate glass doped with high concentration of α-Fe2O3, the crystallization of Fe3(PO4)2 is pronounced. The assignments of IR band positions and the corresponding interpretation are given. The importance of this study for the technology of vitrification of high-level radioactive wastes is emphasized.

Redox control of electric melters with complex feed compositions II. Preliminary limits for radioactive waste melters

The redox state of glass from electric melters with complex feed compositions is determined by balance between gases above the melt, and transition metals and organic compounds in the feed. Part II develops preliminary operating limits for the vitrification of high-level radioactive waste. Limits on reducing potential to preclude the accumulation of combustible gases, accumulation of sulfides and selenides, and degradation of melter components are the most critical. Problems associated with excessively oxidizing conditions, such as glass foaming and potential ruthenium volatility, are controlled when sufficient formic acid is added to adjust melter feed rheology.

Electric melting of nuclear waste glasses State of the art

Vitrification of high-level liquid radioactive wastes has become the international approach for converting highly radioactive wastes into a durable solid. With the exception of France and the first production facility in England, the technology for converting the liquid wastes into glass is similar to “all-electric” melters used in the commercial glass industry. However, the preferred melter feed is a 25–40 wt.% slurry rather than dry batch. The general criteria and bases for design of these melters are presented with a brief review of the historical development. The primary focus of this paper is to review the current status of the designs of the melters in the United States, West Germany, Japan and England. The primary lessons learned in these countries' development programs will be summarized. The influence of melting problems such as corrosion, crystalline sludge, reboil and metal precipitation on the melter designs is reviewed. Future refinements and potential problems that may influence future designs are identified.

A pilot plant demonstration of the vitrification of radioactive solutions using microwave power

A process has been developed that exploits the characteristics of microwave heating for the vitrification of high-level radioactive liquid waste. This process, known as microwave vitrification, has been successfully operated at pilot plant scale in an active cell using simulated liquid waste containing several curies of radioactivity. Excellent decontamination factors have been achieved for both volatiles and nonvolatiles with an average ruthenium decontamination factor of 490 and a gross alpha emitter decontamination factor of 100,000, and almost all the radioactivity being incorporated in a glass block.

Technetium Volatilization During HLLW Vitrification

AbstractThe volatilization of technetium in the course of high-level waste vitrification may be considerable, depending on process variations. A series of experiments has been performed in order to gain more insight into the vaporization behavior of Tc during HLLW vitrification. No losses have been observed during chemical denitration or drying of the solutions on a drum drier at moderate temperatures. However, substantial volatilization (40–60 %) occurred during high temperature processing which applied either liquid feeding or feeding as a dehydrated product into the glass melter. Details of parameters investigated are described, and consequences for technical process application are discussed.

Generation and Vitrification of High-Level Light Water Reactor Liquid Waste

The Nuclear Waste Vitrification Project was conducted to demonstrate the vitrification of high-level liquid waste (HLLW) generated during the reprocessing of spent fuel discharged from an operating light water reactor. Six pressurized water reactor fuel assemblies, containing 2.3 tU, were processed for the generation at HLLW. A conventional Purex-type process was used for the first cycle so that the HLLW generated would be typical of the nitric acid, fission product waste stream from the first extraction cycle of a commercial plant. Uranium and nonradioactive chemicals, normally added to the HLLW by back-cycling of waste from second and third solvent-extraction cycles, were added to the dilute HLLW to produce a waste composition typical of the HLLW from a commercial plant. The waste was then concentrated tenfold to provide feed for solidification by the spray calciner/in-can melting process. During calcination, the liquid waste was pumped to the calciner vessel, which was heated to 750/sup 0/C. The powdered calcine fell into a stainless steel canister, which was maintained at 1050/sup 0/C; this canister was attached directly to the bottom of the calciner. Glass-forming chemicals were metered into the canister simultaneously with the calcine. After the materials melted, the canister was cooled to producemore » glass. Two canisters containing glass were produced.« less

Vitrification of high-level wastes: A review of the computer thermal analyses for storage canisters

CANIST, a two-dimensional ( r and θ) computer program that solves the unsteady-state, heat-conduction equation, was used to model the thermal behavior of canisters filled with waste glass. CANIST has been found to be a valuable analytical tool for predicting the temperature profile of a waste storage canister as a function of several variables, including the diameter of the canister, the placement of internal fins, the heat generation rate of the waste glass, and the thermophysical properties of the canister and the waste glass. Thus, temperature-dependent processes that may affect the integrity of the glass/canister unit, for example cracking, can be investigated using an analytical approach.

Nuclear Waste Vitrification in a Ceramic-Lined Electric Glass Melter

Vitrification of high-level liquid waste from the nuclear fuel cycle in a liquid-fed electric glass melter is described with particular reference to the present status of the melter technology developed at the Institut für Nukleare Entsorgungstechnik (INE), Karlsruhe, West Germany. Application of the process in an active plant to be built at the Eurochemic Site at Mol, Belgium, is outlined. Based on available experience with simulated waste solutions, it is shown that electric melting of waste glass is a promising technique that can substantially simplify nuclear waste vitrification.

Vitrification of Simulated High-Level Radioactive Waste by a Slurry-Fed Ceramic Melter

Savannah River Laboratory is testing the slurry feeding of a ceramic melter as a possible method to vitrify Savannah River Plant high-level radioactive waste. Feeding a liquid slurry requires simpler and less expensive equipment than does feeding a powdered calcine. Experiments have progressed from manual feeding, to a semiautomatic system, to the present slurry-feed system, which is completely automatic. All experiments to date indicate that slurry feeding is a promising way of vitrifying waste. No safety hazards associated with feeding the slurry onto molten glass at 1150°C have been observed experimentally, even when the melter chamber was purposely flooded.

Incineration/vitrification of simulated low-level institutional wastes in a joule-heated glass melter

The purpose of this study was to determine the feasibility of reducing the volume of low-level institutional wastes by incineration and then converting the residual solids into glass with a single-step process. Experimental test results with simulated waste show that a joule-heated ceramic melter originally developed for high-level waste vitrification can be used for the low-level waste volume reduction process. Off gas analysis from the experiment shows that incineration of combustible solids was essentially complete. This is attributable to high combustion temperatures (>950 °C) and the introduction of excess combustion air or oxygen at the melter's glass surface. The only instance where a pyrolysis product (hydrogen) was detected was when the oxygen concentration in the process off gas depleted to 7.8%. An insulating blanket over the glass surface, which would have reduced incineration temperatures, never formed. In addition, calculations show that for scaled-up processing rates, incineration temperatures will remain high because the medicinal waste provides much of its own heat for combustion, vitrification, and process heat losses. Results of the incineration experiment indicate that process scale-up for higher capacities is quite feasible. Although feed rates in this experiment through the pilot-scale melter were low (0.5 kg/ h), they could be increased as much as tenfold, even for the laboratory-scaled unit tested here, by use of an improved feed system. The overall test results were so encouraging that we recommend further tests aimed at adapting high-level waste vitrification technology to the management of both institutional wastes and low-level wastes generated from the nuclear fuel cycle.

Vitrification of High-Level Radioactive Waste in a Small-Scale Joule-Heated Ceramic Melter

ABSTRACTVitrification is the reference process for the immobilization of radioactive waste from the production of defense materials at the Savannah River Plant (SRP). Since 1979, a small vitrification facility (1 Ib/hr) has been operated at the Savannah River Laboratory using actual SRP waste. In previous studies. dried waste was fed to this smaller melter. This report discusses direct feeding of actual liquid-waste slurries to the small melter. These liquidfeeding tests demonstrated that addition of premelted glass frit to the waste slurry reduces the amount of material volatilized. Results of these tests are in accord with results of large-scale tests with actual waste.

Mercury Reduction and Removal During High-Level Radioactive Waste Processing and Vitrification

ABSTRACTA reference process for immobilizing the high-level radioactive waste in borosilicate glass has been developed at the Savannah River Plant. This waste contains a substantial amount of mercury from separations processing. Because mercury will not remain in borosilicate glass at the processing temperature, mercury must be removed before vitrification or must be handled in the off-gas system. A process has been developed to remove mercury by reduction with formic acid prior to vitrification. Additional benefits of formic acid treatment include improved sludge handling and glass melter redox control.