Abstract: There is much concern that long-lived radioactive elements incorporated in glasses for the disposal of highly active nuclear waste might eventually return to the environment. The most important mechanism begins with leaching of the glass by groundwater and a leach rate, usually based on laboratory tests of simulated vitrified radioactive waste, has been used as the basis of analyses of the radiological safety of a waste repository1–3. The possibility that radiation damage to the glass from the products of nuclear decay could change the leach rate has been studied both by the incorporation of active elements in the glass and by irradiation from external sources. The most important contribution to radiation damage comes from the recoil nuclei during α decay4. Typically the recoil nucleus has a kinetic energy of ∼100 keV and displaces >1,000 atoms in the glass. Experiments which simulate this damage by incorporating short-lived α-emitting isotopes such as 238Pu into the glass have not shown significant increases in leach rate at doses equivalent to >1018 α decays g−1 (refs 5–9). Figure 1 shows that for wastes considered in the UK this corresponds to a time after vitrification of at least 103–104 yr. Recent work with ion beams inducing the radiation damage10–12 has suggested that large increases (up to a factor of just over 50) in leach rate will occur after a critical dose of radiation from the α-decay processes. In addition, experiments in which the leaching solution is irradiated with γ rays13–15 have shown that radiolysis effects in the water can increase the leach rates of glasses. We present here new data and calculations which relate these studies to the conditions expected in a real waste repository. Our analysis is described in more detail elsewhere16,17.

Abstract: Leaching of waste glass in lead and aluminum containers is much less extreme than in relatively inert Teflon containers. Lead and aluminum are already included in several waste-package designs. The chemical durability imparted to a waste glass by sorption of lead and aluminum corrosion products onto the glass surfaces, as reported in this study, is an additional reason to include these metals in the waste-package assembly. Glass corrosion was inhibited in lead containers even though the solution pH rose to values considerably higher than necessary to accelerate silica dissolution in the absence of lead. Relatively minor effects on glass leaching were found in copper, tin, and titanium containers when compared to those measured in Teflon containers. No sorption of materials from any of these containers onto the glasses was evident, although sorption of glass corrosion products onto the container walls was observed.

Abstract: Waste glasses of different compositions were compared in terms of leachability, viscosity, liquidus temperature, and coefficient of expansion. The compositions of the glasses were determined by statistical optimization. Waste glass of the optimized composition is more durable than the current reference composition but can still be processed at low temperature.

Abstract: Surface layers are a common feature of leached surfaces of borosilicate waste glasses. The question of whether these layers can protect the glass against further attack by decreasing the leach rate is still a subject of controversy. The present study investigates the effects of layer formation on leaching kinetics of a borosilicate waste glass containing 20 wt % LWR-type simulated waste oxides: (1) surface layer formation on the borosilicate waste glass in a closed system can be divided in a sequence of overlapping processes. The formation of amorphous phases on the glass surface is observed first, followed by crystalline phase formation, with new phases still appearing after one year in NaCl solution; (2) the complexity of the kinetics is reflected in the variation of concentration of different elements in solution; (3) the overall leaching process varies as a function of time with t/sup n/ and n < 0.3. This reflects that the process cannot be described in terms of either diffusion (n = 0.5) or linear dissolution (n = 1), or a combination of both; and (4) though the details of the process are not yet understood, it is possible to extract a parameter, n, to describe the total massmore » release from the glass in the longer term in NaCl solution if leaching data are measured for at least one year.« less

Abstract: 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 fur 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.

Abstract: The possibility of converting high-level wastes (HLW) to glass was first pursued in Canada and England at a time when other countries were evaluating many other alternatives. By 1966, the British had completed radioactive demonstrations of the FINGAL pot process, converting HLW to borosilicate glass. By this time other countries, including France and the United States, had begun using the glass waste form. Beginning in 1966, several processes, including phosphate and borosilicate glass, were demonstrated by the US in the Waste Solidification Engineering Prototypes (WSEP) program at the Pacific Northwest Laboratory (PNL). Most of the current vitrification processes are adaptations of the FINGAL pot process or the continuous metallic melter used in the WSEP program. One notable exception is the joule-heated ceramic melter, which was adapted from commercial glass technology for HLW by PNL in the mid-1970's. Both batch and continuous processes have been developed to an advanced stage of readiness. These processes are described and compared in this paper.

Abstract: The in situ vitrification process builds upon the electric melter technology previously developed for high-level waste immobilization In situ vitrification converts buried wastes and contaminated soil to an extremely durable glass and crystalline waste form by melting the materials, in place, using joule heating Once the waste materials have been solidified, the high integrity waste form should not cause future ground subsidence Environmental transport of the waste due to water or wind erosion, and plant or animal intrusion, is minimized Environmental studies are currently being conducted to determine whether additional stabilization is required for certain in-ground transuranic waste sites An applications analysis has been performed to identify several in situ vitrification process limitations which may exist at transuranic waste sites Based on the process limit analysis, in situ vitrification is well suited for solidification of most in-ground transuranic wastes The process is best suited for liquid disposal sites A site-specific performance analysis, based on safety, health, environmental, and economic assessments, will be required to determine for which sites in situ vitrification is an acceptable disposal technique Process economics of in situ vitrification compare favorably with other in-situ solidification processes and are an order of magnitude less than the costs formore » exhumation and disposal in a repository Leachability of the vitrified product compares closely with that of Pyrex glass and is significantly better than granite, marble, or bottle glass Total release to the environment from a vitrified waste site is estimated to be less than 10/sup -5/ parts per year 32 figures, 30 tables« less

Abstract: 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

Abstract: PURPOSE:To easily form an extremely thin glass coating without any pinholes by heating the particle film consisting essentially of silicic acid obtained by supplying a glass type raw material on a metallic surface and causing oxidation reaction in flames in a temp. region at which said metal is not melted. CONSTITUTION:Mainly a wire-or strip-like metal is fed into a sealed vessel made of quartz having a gas supply system and a gas exhaust system, and gaseous mixture consisting essentially of SiCl4, H2, O2 are flame hydrolyzed in this vessel. As result, the particle film consisting essentially of SiO2 which is the product of reaction is formed on the surface of said metal. Thence, this particle film is heated usually at temps. around 1,500 deg.C at which the metal is not melted, by a ''Siliconit '' furnace, furnace, a high frequency furnace or an oxygen burner, whereby it is turned to transparent glass and a quartz type glass film of <=several mum is formed. If the m.p. of the metal is below the vitrification temp., it is a good practice to heat only the surface thereof to high temp. instantaneously. The metal above the vitrification temp. includes, for example, Pt, and Rh, whilist the metal below the vitrification temp. includes, for example copper, iron, etc.

Abstract: A new process for stabilizing buried radioactive wastes without exhumation is being developed by Pacific Northwest Laboratory (PNL). The process, known as in situ vitrification, converts waste and contaminated soil to a durable glass and crystalline material by passing an electric current between electrodes placed in the ground. Joule heating created by the flowing current has generated temperatures over 1700/sup 0/C which cause the soil to melt and dissolve or encapsulate the wastes. Engineering-scale tests conducted in the laboratory have melted approximately 45 kgs (30 liters) of soil at a time by this technique. Encouraging results from these engineering-scale tests led to the design and construction of a pilot-scale field test unit which has solidified approximately 9000 kg of simulated contaminated soil per test. Test results and evaluations to date have been very promising. No detectable migration of hazardous species into uncontaminated soil has been found, and volatilization during melting has been very low. Leach studies have found the vitrified soil to be a highly durable waste form similar to pyrex glass. Electrical power costs to solidify a disposal site have been calculated at less than $70 per cubic meter ($2/ft/sup 3/) of waste. Future activities include both radioactive and more » nonradioactive pilot and large-scale tests. « less

Abstract: Silicon cells are provided with electrical conductor wires and then, before they are assembled in modules, each is dipped in a glass slurry with a low viscosity to obtain a coating. The coating is vitrified by heat to form a protective glass capsule adhering to both the Si and the wires. The slurry pref. uses a glass powder contg. a metal oxide, esp. V2O5, which lowers the transformation temp. of the glass without appreciably increasing its coefft. of thermal expansion. The slurry pref. also contains water and a clay providing thixotropy; or an organic solvent and binder, esp. terpineol plus ethyl cellulose. A glass is selected with an elastic modulus and a coefft. of expansion similar to Si, and a vitrification temp. not damaging the cell. The glass has good transparency and resists moisture; it can form an anti-reflection coating due to its refractive index. In conventional mfg. methods, the completed module is encapsulated. The invention uses the encapsulation of the individual cells prior to assembly in a module so mfg. costs are decreased.

Abstract: A small-scale joule-heated ceramic melter contained in the Shielded Cells Facility has demonstrated the vitrification process for actual Savannah River Plant radioactive waste. Losses of radionuclides due to volatility are low and easily treated, and the glass produced is of comparable quality to laboratory-prepared simulated glass. Future work will include studies with wastes from other tanks, using new frit compositions. Leaching tests will continue, with emphasis being placed on long-term tests under anticipated repository conditions.

Abstract: Radioactive defense waste currently stored at the Savannah River Plant in liquid form is to be immobilized by incorporation into a borosilicate glass. The glass melter for this process will consist of a refractory lined tank enclosed in a water-cooled, steel vessel and fed by an aqueous slurry of glass frit plus radioactive waste. As an aid to understanding the melting process and scaling data from small melters, conservation principles were used to develop equations for predicting the melt rate for a feedpile of arbitrary radius plus additional information regarding feedpile behavior and energy consumption.

Abstract: .....-------DIICLAIMIII -------. TMt rt00t1 wtt � • ., ICC'Ou !'ll of woriiiiPC!MDftd bv ¥1 IIII'I"'CY ot rt'4 Un111d l!twt Gt�WmMin\. N•lthtt lhl UnH«i St.ta Gcwtrnmtn t not' any Jlfi'ICY thtf.of, nor .,yo! thllr ""Piov•. M'ltkftM'IY wtHII!nfy, I•P'tftl df 1�1«1;, Qf INIUII'M any I.,.C ll«<;lllty Of ffti)OI'IIIbll!ty IOf llilt tttUfliCV, tol"'pl tfiMU, 01 "•h.ll'ntt� of •nv lnlotrNifO!'I, eppetfNI, PfQd\ICl, Of l)f� dltdoM, � tiJ)r.w'lll tf'll'! Itt 1.1• w.cvld 1'\Qi lnfrl,.,_. �tlw!tly owniQ rltM:t. A.� Mrtll'l to anv �p«lf4c comi'TW�:I.et prodvd, pr� OJ twvlw by tt.d• Nlmt, tttct.mtf\:, f!'llti'N ftttul'lf, Of othtt'wt•, oa.. not n«:fttllr l1y etll'lltllwtl or Imply iu fndo#'•ment, rKOmmendt!tot'l, Of flYOI'Ing by th41 Unltlld ltt:tft OoWfrvrwu r.w •rw IQII"'C¥' thtt.af. 1M vltv.1 tf!Cf oplnlom ot Nthorl u:prttMd -.. n do !"'lt f*ftMri!y tlttl 01 r.tltct tho• ol 1"U1111.:! SlfltJ QQWI'nm.nt 0!' .tnv ti',Jltni:V thtrtof,

Abstract: Pacific Northwest Laboratory has completed a s eries of experimental tests sponsored by the U.S. Department of Energy (DOE) to determine the feasibility of incinerating and vitrifying organic ion-exchange resins in a single-step process. The resins used in this study were identical to those used for decon­ taminating auxiliary building water at the Three Mile Island ( TMI) Unit 2 reac­ tor. The primarily organic resins were loaded with nonradioactive isotopes of cesium and strontium for processing in a pilot-scale, joule-heated glass melter modified to support resin combustion. The glass melter contains molten glass which is electrically heated. The resins are fed to the melter via a novel technique, which forces contact of the combustion gases and molten glass. The glass, which absorbs cesium and stron­ tium from the combustion gases, is poured and cooled in a disposal container. Thus the cesium and strontium from the loaded resin are incorporated in a chemically stable glass product. The feasibility of the vitrification process for organic resins is determined by various parameters. These include: • process rates • v o 1 ume reduc.t ion • radionuclide retention in the glass • material conform ance to operating conditions • process stability. · The feasibility tests demonstrated an average process rate of 3.0 kg/h. Based on this rate, if 50 organic resin liners were vitrified in a six-month campaign, a melter 2. 5 times the size of the pilot scale unit would be adequate. A maximum achievable volume reduction of 91% was demonstrated in these tests. This means that the glass produced from the melter was 9% of the original vol­ ume of the resins. The unique feeding technique was successful in absorbing most of the radionuclides into the glass. The average cesium retention in the glass throughout the feasibility tests was 84%; strontium retention was held at 95%. The portion of radionuclides entrained with the gaseous effluents may

Abstract: PURPOSE:To manufacture a glass base material for an optical fiber with fluoride glass having a small vitrification tendency by heating a glass rod for a core to a temp. close to the glass transition temp. and depositing molten glass for a clad on the surface of the heated rod in the form of a layer. CONSTITUTION:When a glass base material for an optical fiber having a waveguide structure is manufactured with fluoride glass having a small vitrification tendency, while rotating a glass rod 15 for a core with chucks 4, the rod 15 is uniformly heated to a temp. close to the glass transition temp. with a heating source 17 and kept at the temp. Glass powder for a clad is then heated and melted with a melting device 18 while moving a feeder 16 over the full length of the rod 15, and the molten glass is uniformly deposited on the surface of the rod 15 in the form of a layer and rapidly cooled. Since a glass layer for a clad can be formed on the surface of the rod 15 without causing crystallization, a glass base material for an optical fiber having a waveguide structure can be easily manufactured with fluoride glass requiring rapid cooling for vitrification.

Abstract: Heating by microwave energy is explored in processing of two nuclear waste forms: 1)drying of a pelleted form and 2)vitrification. It is shown that residence time can be greatly reduced compared to conventional heating sources, without affecting product quality.

Abstract: The vitrification of simulated nuclear waste calcines was studied in a ceramic-lined melter with a glass surface area of 0.76 m2. The melter contained 0.25 m3 of glass heated by the flow of an AC current (ranging from 600-1200 A) between two Inconel-690 slab-type electrodes immersed directly in the glass at either end of the melter tank. The melter was maintained at operating temperatures for 13.5 months and produced 62000 kg of glass. The maximum sustained operating period was 122 h, during which glass was produced at the rate of 70 kg/h. The basis design features of the melter, and some of the operating experiences, are discussed. The proposed use of ceramic-lined electric-heated glass melters for the vitrification of nuclear wastes is feasible, but additional studies of the chemistry of the melting process and of the corrosion process of the complex nuclear waste glasses are indicated.

Abstract: Schemes to dispose of high-level radioactive waste by vitrification and subsequent burial call for an appraisal of the consequences of groundwater encountering the waste glass1. Glass dissolution rates are influenced not only by properties of the glass and solvent themselves but also by properties of the combination such as the glass surface area to solvent volume ratio2. Although pressure may affect glass dissolution rates, it has not received unambiguous explicit experimental investigation. A simple experiment has been designed to determine dissolution rates at room temperature of glass samples subject to very high pressure in a centrifuge. The results reported here show that the dissolution rates for Na and Si increase slightly with pressure, but that at a pressure of 160 MPa (>1,500 atm) these dissolution rates are still of the same order as those at atmospheric pressure.

Abstract: Pacific Northwest Laboratory (PNL) is demonstrating a vitrification system designed for immobilization of highly loaded SDS zeolites. The Zeolite Vitrification Demonstration Project (ZVDP) utilizes an in-can melting process. All steps of the process have been demonstrated, from receipt of the liners through characterization of the vitrified product. The system has been tested with both nonradioactive and radioactive zeolite material. Additional high-radioactivity demonstrations are scheduled to begin in FY-83. 5 figures, 4 tables.

Abstract: The effects of solidification pressure and cooling rate on the as-prepared density of polystyrene have been determined simultaneously. In general, the dependence of density individually on vitrification pressure and cooling rate is as previously reported; i.e., density decreases with increasing cooling rate and increases in a logarithmic fashion with pressure. However, we observe that the variation of density with vitrification pressure (∂p ρ/∂Pv) is independent of cooling rate; that the variation of density with vitrification cooling rate (∂ρ/∂Rv) is independent of pressure; and that the as-prepared density varies linearly with log(vitrification cooling rate).

Abstract: In order to produce doped SiO2 glass for glass fibres, quartz powder or fine particles of SiO2 glass are treated with a gas containing SiCl4, a gaseous additive and steam in order to add the dopant to the glass element This glass element is melted at high temperature to produce transparent, doped SiO2 glass The production of the glass particles, the addition of the dopant and the vitrification of the glass element are carried out in separate process steps, in each case under suitable conditions The production rate is significantly increased due to the individual process steps The dopant content is unlimited and can be adjusted to any desired value by changing the reaction time during dissolution It is also possible to add dopants such as PbO2, SnO2 and ZnO which are otherwise extremely difficult to employ to the glass element In order to produce the glass fibre preform, the doped SiO2 glass is precipitated and melted on a starting material which is tilted at an angle of from 5 to 90 DEG with respect to the beam direction of the doped SiO2 glass This gives a transparent, doped SiO2 glass element of uniform external diameter and uniform boundary surface at a high synthesis rate This enables the mass production of glass fibres at low cost