Abstract: A method of vitrifying soil at or below a soil surface location. Two or more conductive electrodes are inserted into the soil for heating of the soil mass between them to a temperature above its melting temperature. Materials in the soil, such as buried waste, can thereby be effectively immobilized.

Abstract: Glass has been widely advocated as a suitable medium for the immobilization of high-level nuclear waste1–6. Methods of vitrification to borosilicate glass are advanced, with processes set up on a semi-industrial scale7–9, but an alternative strategy would be to incorporate waste into a high-silica glass10. It is generally proposed5,6 that vitrified radioactive waste be enclosed in metal canisters and stored underground. However, the presence of heated groundwater in rocks means that the canisters may corrode, allowing hot aqueous fluids to come into contact with the radioactive waste glass and leach out radioactive elements. Numerous laboratory tests have been performed for short periods and at relatively low temperatures to assess the leaching performances of different types of glass11–15, however, extrapolation to predict the long-term behaviour of glasses after burial is very uncertain. I report here a microprobe analysis technique which investigated hydrothermal leaching of rhyolite glass adjacent to a fluid conduit in the Tertiary hydrothermal system of the Isle of Skye, north-west Scotland. As the composition of rhyolite and proposed high-silica radioactive waste glasses are similar, this study may help to predict the long-term leaching behaviour of such glasses after underground burial.

Abstract: Method for the ultimate disposition of radioactive wastes by vitrification, in which weak to medium radioactive waste concentrates from borate-containing radioactive liquids are mixed with added glass-forming materials, maximally in a ratio of 1:3, and the mixture heated to obtain a glass-forming melt

Abstract: In the vitrification of HLW the aim is to create a product, the stability (thermal, mechanical) of which is, to the greatest possible degree, independent of fluctuations in its composition. Process control during vitrification of HLLW is necessary in order to assure the quality of the final glass product. The chemical processes are controlled by quantitative analysis of all material streams entering and leaving the melting device. There are six material streams to consider: frit input, waste input, corrosion products (generated by the vitrification process), off-gas, recycled off-gas, glass output. Additionally, the capacity of the glass for buffering short-term fluctuations in the incoming streams must be known. Sufficient sampling during non-radioactive vitrification in addition to materials tests can provide the necessary data basis. The results are combined in a materials balance.

Abstract: PURPOSE:To inexpensively obtain a high quality optical fiber by converting glass starting material gases into glass soot in a flame cone to form porous glass on the tip of a starting member and transparently vitrifying the glass by the heat of the cone. CONSTITUTION:Glass starting materials fed into oxyhydrogen burner 21 cause oxidation reaction in flame 22, generating glass soot. In the early stage this soot is blown on the tip of starting member 5 entering the inside of flame cone 23, and it is deposited to form porous glass 25. By pulling up member 5 in synchronism with the forming speed, glass 25 is passed through cone 23 which is a very high temp. region. At this time, glass 25 is freed of bubbles and made transparent. It is fed upward and then converted into optical fiber base material 26 of transparent glass. Thus, by applying the heat source for causing the glass synthesis reaction to the transparent vitrification, a high quality optical fiber is manufactured at a low cost.

Abstract: The design of a slurry-fed electric gas melter and an examination of the performance and condition of the construction materials were completed. The joule-heated, ceramic-lined melter was constructed to test the applicability of materials and processes for high-level waste vitrification. The developmental Liquid-Fed Ceramic Melter (LFCM) was operated for three years with simulated high-level waste and was subjected to conditions more severe than those expected for a nuclear waste vitrification plant.

Abstract: During FY-1980, Pacific Northwest Laboratory (PNL) tested three vitrification processes on simulated high-level radioactive waste typical of that stored or being produced at US defense facilities. Processes tested included a spray calciner/in-can melter, spray calciner/ceramic melter and direct liquid feeding of a ceramic melter. Tests were made on pilot-scale as well as fullscale equipment. Over 16,000 kg of glass product were produced from 68,000 L of simulated waste. Several compositions were tested, and the glass products were evaluated. Emphasis was placed on determining the processing rates and the ability of the waste to be processed. Off-gas data were collected on several runs. Major conclusions drawn from this test program are divided into processing results, glass-product results, and general information.

Abstract: The conformational and enthalpic changes that occur in poly(vinyl chloride) (PVC) glasses that have been vitrified from the melt under pressure have been examined by Fourier transform infrared spectroscopy and quantitative differential scanning calorimetry. It is shown that these pressures freeze in the high energy states that are characteristic of the vitrification temperature and increase the apparent glass transition temperature of the polymer. In addition, pressures in excess of the vitrification pressure, cause intermolecular effects that can be relaxed out below Tg. Both of these processes create characteristic endothermic and exothermic changes in the apparent heat capacity of the glass that appear over a period of time and are sensitive functions of the glass formation processes as well as the subsequent annealing history. The endothermic events are interpreted as the stress perturbed volumetric relaxation process white the exotherms are associated withh the release of the frozen in stresses.

Abstract: A 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.

Abstract: Current plans indicate that the high-level wastes stored at the Savannah River Plant will be solidified by vitrification. The behavior of mercury and iodine during the vitrification process is of concern because: mercury is present in the waste in high concentrations (0.1 to 2.8 wt%); mercury will react with iodine and the other halogens present in the waste during vitrification and; the mercury compounds formed will be volatilized from the vitrification process placing a high particulate load in the vitrification system off-gas. Twelve experiments were completed to study the behavior of mercury during vitrification of simulated SRP Purex waste. The mercury was completely volatized from the vitrification system in all experiments. The mercury reacted with iodine, chlorine and oxygen to form a fine particulate solid. Quantitative recovery of mercury compounds formed in the vitrification system off-gas was not possible due to high (37 to 90%) deposition of solids in the off-gas piping. The behavior of mercury and iodine was most strongly influenced by the vitrification system atmosphere. During experiments performed in which the oxygen content of the vitrification system atmosphere was low (< 1 vol%); iodine retention in the glass product was 27 to 55%, the mercury composition of themore » solids recovered from the off-gas scrub solutions was 75 to 85 wt%, and a small quantity of metallic mercury was recovered from the off-gas scrub solution. During experiments performed in which the oxygen content of the vitrification system atmosphere was high (20 vol%), iodide retention in the glass product was 3 to 15%, the mercury composition of the solids recovered from the off-gas scrub solutions was 60 to 80 wt%, and very little metallic mercury was recovered from the off-gas scrub solution.« less

Abstract: PURPOSE:To obtain the titled base material without exerting unfavorable influence on glass fine powder and transparent glass by heat vitrifying glass material subjected to combustion oxidation at a temp. near the transparently vitrifying temp. while feeding O2 gas into a pipelike starting member. CONSTITUTION:Glass fine powdered body 22 is heated ro the transparently vitrifying temp. (about 1,300 deg.C) with heater 28, and while feeding O2 gas into support pipe 23 and substrate 21 from inlet 24, pipe 23 is slowly lowered with lift 26. Body 22 is successively converted into transparent glass 29 from the lower end. At the same time, substrate 21 is converted into B2O3 by combustion vitrification, and it is partially evaporated. When the whole of body 22 is converted into transparent glass 29, starting member 21 disppears thoroughly or remains as glass. No stress is imposed on glass 29 from member 21, and the remaining glass may be removed with hydrofluoric acid.

Abstract: Vitrification 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 lb/h) 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 liquid-feeding 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.

Abstract: Three full-scale vitrification processes at the Pacific Northwest Laboratory produced over 67,000 kg of simulated nuclear-waste glass from March 1979 to August 1980. Samples were analyzed to monitor process operation and evaluate the resulting glass product. These processes are: Spray Calciner/In-Can Melter (SC/ICM); Spray Calciner/Calcine-Fed Ceramic Melter (SC/CFCM); and Liquid-Fed Ceramic Melter (LFCM). Waste components in the process feed varied less than +- 10%. The SC/ICM and SC/CFCM which use separate waste and frit feed systems showed larger glass compositional variation than the LFCM, which processed only premixed feed during this period. The SC/ICM and SC/CFCM product contained significant amounts of acmite crystals, while the LFCM product was largely amorphous. In addition, the lower portion of all SC/ICM-filled canisters contained a zone rich in waste components. A product chemical durability as determined by pH4 and soxhlet leach tests varied considerably. Aside from increased durability under pH4 conditions with decreasing waste content, glass composition, microstructure and melting process did not correlate with glass durability. For all samples analyzed, the weight loss under pH4 conditions ranged from 17.7 to 85.2 wt %. Soxhlet conditions produced weight losses from 1.78 to 3.56 wt %.

Abstract: At the Savannah River Plant, the reference process for the immobilization of defense high-level waste (DHLW) for geologic storage is vitrification into borosilicate glass. During geologic storage for 106y, the glass would be exposed to ∼3 × 1010 rad of β radiation, ∼1010 rad of γ radiation, and 1018 particles/g glass for both α and α-recoil radiation. This paper discusses tests of the effect of these radiations on the leachability and density of the glass. No effect of the radiations was detected that reduced the effectiveness of the glass for long-term storage of DHLW even at doses corresponding to 106 years storage for the actual glass. For the tests, glass containing simulated DHLW was prepared from frit of the reference composition. Three methods were used to irradiate the glass: external irradiations with beams of ∼200 keV or Pb ions, internal irradiations with Cm–244 doped glass, and external irradiations with Co–60 γ rays. Results with both Xe and Pb ions indicate that a dose of 3 × 1013 ions/cm2 (simulating >106 years storage) does not significantly increase the leachability of the glass in deionized water. Tests with Cm–244 doped glass show no increase in leach rate in deionized water up to a dose of 1.3 × 1018 α and α-recoils/g glass. The density of the Cm–244 doped glass has decreased by 1% at a dose of 1018 particles/g glass. With γ-radiation, the density has changed by <0.05% at a dose of 8.5 × 1010 rad. Results of leach tests in deionized water and brine indicated that this very large dose of γ-radiation increased the leach rate by only 20%. Also, the leach rates are 3 to 4 times lower in brine.