water alteration processes and kinetics of basaltic glasses, natural analogue of nuclear glasses

International Nuclear Information System (INIS)

Techer, I.; Advocat, Th.; Vernaz, E.; Lancelot, J.R.; Liotard, J.M.

1997-01-01

Dissolution experiments of a basaltic glass were carried out at 90 deg C for different reaction progresses. The initial dissolution rate was compared with values obtained for rhyolitic glass and the R7T7 nuclear glass. The activation energy was also determined by computing literature data. The results provide similar reactional mechanism for basaltic and nuclear glasses. Dissolution rates measured under saturation conditions were compared to theoretical dissolution rates. These ones were calculated using two kinetic models: the first rate equation is the Grambow's law which only takes into account ortho-silica acid activity; the second rate equation was proposed by Daux et al., where silica and aluminum are combined to formulate the affinity. The comparison between experimental and theoretical results point out that these two models are not appropriate to describe the alteration kinetic of basaltic glasses. (authors)

Application of ceramic and glass materials in nuclear power plants

International Nuclear Information System (INIS)

Hamnabard, Z.

2008-01-01

Ceramic and glass are high temperature materials that can be used in many fields of application in nuclear industries. First, it is known that nuclear fuel UO 2 is a ceramic material. Also, ability to absorb neutrons without forming long lived radio-nuclides make the non-oxide ceramics attractive as an absorbent for neutron radiation arising in nuclear power plants. Glass-ceramic materials are a new type of ceramic that produced by the controlled nucleation and crystallization of glass, and have several advantages such as very low or null porosity, uniformity of microstructure, high chemical resistance etc. over conventional powder processed ceramics. These ceramic materials are synthesized in different systems based on their properties and applications. In nuclear industries, those are resistant to leaching and radiation damage for thousands of years, Such as glass-ceramics designed for radioactive waste immobilization and machinable glass-ceramics are used. This article introduces requirements of different glass and ceramic materials used in nuclear power plants and have been focused on developments in properties and application of them

Thermodynamic model of natural, medieval and nuclear waste glass durability

International Nuclear Information System (INIS)

Jantzen, C.M.; Plodinec, M.J.

1983-01-01

A thermodynamic model of glass durability based on hydration of structural units has been applied to natural glass, medieval window glasses, and glasses containing nuclear waste. The relative durability predicted from the calculated thermodynamics correlates directly with the experimentally observed release of structural silicon in the leaching solution in short-term laboratory tests. By choosing natural glasses and ancient glasses whose long-term performance is known, and which bracket the durability of waste glasses, the long-term stability of nuclear waste glasses can be interpolated among these materials. The current Savannah River defense waste glass formulation is as durable as natural basalt from the Hanford Reservation (10 6 years old). The thermodynamic hydration energy is shown to be related to the bond energetics of the glass. 69 references, 2 figures, 1 table

Structural role of molybdenum in nuclear glasses: an EXAFS study

International Nuclear Information System (INIS)

Calas, G.; Le Grand, M.; Galoisy, L.; Ghaleb, D.

2003-01-01

The Mo environment has been investigated in inactive nuclear glasses using extended X-ray absorption spectroscopy (XAS). Mo is present in a tetrahedron coordinated to oxygen in the form of molybdate groups [MoO 4 ] 2- (d(Mo-O)=1.78 A). This surrounding is not affected by the presence of noble metal phases in the nuclear glass. Relying on the XAS results, on the bond-valence model and on molecular dynamics simulations of a simplified borosilicate model glass, we show that these groups are not directly linked to the borosilicate network but rather located within alkali and alkaline-earth rich domains in the glass. This specific location in the glass network is a way to understand the low solubility of Mo in glasses melted under oxidizing conditions. It also explains the possible phase separation of a yellow phase enriched in alkali molybdates in molten nuclear glasses or the nucleation of calcium molybdates during thermal aging of these glasses. Boron coordination changes in the molten and the glassy states may explain the difference in the composition of the crystalline molybdates, as they exert a direct influence on the activity of alkalis in borosilicate glasses and melts

Turning nuclear waste into glass

Energy Technology Data Exchange (ETDEWEB)

Pegg, Ian L.

2015-02-15

Vitrification has emerged as the treatment option of choice for the most dangerous radioactive waste. But dealing with the nuclear waste legacy of the Cold War will require state-of-the-art facilities and advanced glass formulations.

3 and 4 oxidation state element solubilities in borosilicate glasses. Implement to actinides in nuclear glasses; Solubilite des elements aux degres d'oxydation (3) et (4) dans les verres de borosilicate. Application aux actinides dans les verres nucleaires

Energy Technology Data Exchange (ETDEWEB)

Cachia, J.N

2005-12-15

In order to ensure optimal radionuclides containment, the knowledge of the actinide loading limits in nuclear waste glasses and also the comprehension of the solubilization mechanisms of these elements are essential. A first part of this manuscript deals with the study of the differences in solubility of the tri and tetravalent elements (actinides and surrogates) particularly in function of the melting temperature. The results obtained indicate that trivalent elements (La, Gd, Nd, Am, Cm) exhibit a higher solubility than tetravalent elements (Hf, Th, Pu). Consequently, it was planned to reduce plutonium at the oxidation state (III), the later being essentially tetravalent in borosilicate glasses. An innovating reduction process of multi-valent elements (cerium, plutonium) using silicon nitride has been developed in a second part of this work. Reduced plutonium-bearing glasses synthesized by Si{sub 3}N{sub 4} addition made it possible to double the plutonium solubility from 2 to 4 wt% at 1200 deg C. A structural approach to investigate the differences between tri and tetravalent elements was finally undertaken. These investigations were carried out by X-rays Absorption Spectroscopy (EXAFS) and NMR. Trivalent rare earth and actinide elements seem to behave as network modifiers while tetravalent elements rather present true intermediaries' behaviour. (author)

Neural network analysis of nuclear waste glass composition vs durability

International Nuclear Information System (INIS)

Seibel, C.K.

1994-01-01

The relationship between the chemical composition of oxide glasses and their physical properties is poorly understood, but it is becoming more important as vitrification (transformation into glass) of high-level nuclear waste becomes the favored method for long-term storage. The vitrified waste will be stored deep in geologic repositories where it must remain intact for at least 10,000 years. A strong resistance to groundwater exposure; i.c. a slow rate of glass dissolution, is of great importance. This project deals specifically with glass samples developed and tested for the nuclear fuel reprocessing facility near West Valley, New York. This facility needs to dispose of approximately 2.2 million liters of high-level radioactive liquid waste currently stored in stainless steel tanks. A self-organizing, artificial neural network was used to analyze the trends in the glass dissolution data for the effects of composition and the resulting durability of borosilicate glasses in an aqueous environment. This durability data can be used to systematically optimize the properties of the complex nuclear glasses and slow the dissolution rate of radionuclides into the environment

NUCLEAR WASTE GLASSES: CONTINUOUS MELTING AND BULK VITRIFICAITON

International Nuclear Information System (INIS)

KRUGER, A.A.

2008-01-01

This contribution addresses various aspects of nuclear waste vitrification. Nuclear wastes have a variety of components and composition ranges. For each waste composition, the glass must be formulated to possess acceptable processing and product behavior defined in terms of physical and chemical properties that guarantee the glass can be easily made and resist environmental degradation. Glass formulation is facilitated by developing property-composition models, and the strategy of model development and application is reviewed. However, the large variability of waste compositions presents numerous additional challenges: insoluble solids and molten salts may segregate; foam may hinder heat transfer and slow down the process; molten salts may accumulate in container refractory walls; the glass on cooling may precipitate crystalline phases. These problems need targeted exploratory research. Examples of specific problems and their possible solutions are discussed

Radiation and Thermal Ageing of Nuclear Waste Glass

Energy Technology Data Exchange (ETDEWEB)

Weber, William J [ORNL

2014-01-01

The radioactive decay of fission products and actinides incorporated into nuclear waste glass leads to self-heating and self-radiation effects that may affect the stability, structure and performance of the glass in a closed system. Short-lived fission products cause significant self-heating for the first 600 years. Alpha decay of the actinides leads to self-radiation damage that can be significant after a few hundred years, and over the long time periods of geologic disposal, the accumulation of helium and radiation damage from alpha decay may lead to swelling, microstructural evolution and changes in mechanical properties. Four decades of research on the behavior of nuclear waste glass are reviewed.

Stimulated nuclear spin echos and spectral diffusion in glasses

International Nuclear Information System (INIS)

Borges, N.M.; Engelsberg, M.

1984-01-01

Experimental results of stimulated nuclear spin echos decay in glasses are presented. The measurements were performed in B 2 O 3 glasses, at the 23Na and 11 B resonance lines. The data analysis allows the study of Spectral diffusion at an inhomogeneous nuclear magnetic (NMR) resonance line, broadened for a desordered system of nuclear spins. A model is proposed to explain the time constants, and the particular form of the decay. (A.C.A.S.) [pt

Simulation used to qualify nuclear waste glass for disposal

International Nuclear Information System (INIS)

Reimus, T.W.; Kuhn, W.L.

1987-07-01

A hypothetical vitrification system was simulated errors associated with controlling and predicting the composition of the nuclear waste glass produced in the system. The composition of the glass must fall within certain limits to qualify for permanent geologic disposal. The estimated error in predicting the concentrations of various constituents in the glass was 2% to 8%, depending on the strategy for sampling and analyzing the feed and on the assumed magnitudes of the process uncertainties. The estimated error in controlling the glass composition was 2% to 9%, depending on the strategy for sampling and analyzing the waste and on the assumed magnitudes of the uncertainties. This work demonstrates that simulation techniques can be used to assist in qualifying nuclear waste glass for disposal. 3 refs., 2 figs., 4 tabs

Aqueous corrosion of silicate glasses. Analogy between volcanic glasses and the French nuclear waste glass R7T7

International Nuclear Information System (INIS)

Goldschmidt, F.

1991-01-01

The behaviour of borosilicate glasses upon aqueous corrosion is controlled for long periods of time (>10,000 years) by processes which are not directly accessible by means of laboratory experiments. The analogical approach consists here to compare leaching performances between the french nuclear waste glass R7T7 and natural volcanic glasses, basaltic and rhyolitic ones. The three glasses were leached in the same conditions; open system, 90 deg C, initial pH of 9.7. Basaltic and R7T7 glasses having the same kinetic of dissolution, the basaltic glass was chosen as the best analogue. (author). refs., figs., tabs

Investigation of metastable immiscibility in nuclear-waste-glasses. I-III

International Nuclear Information System (INIS)

Egnell, J.; Larsen, J.G.; Moeller, L.; Roed, G.

1981-12-01

Metastable liquid-liquid separation in glasses can often cause significant changes in physical and chemical properties of the original homogeneous glass. In some technical borosilicate glasses this phenomenon is used to change the chemical durability of the glass. For potential nuclear-waste-glasses the slow cooling through the temperature range 550 0 C - 700 0 C may lead to such a liquid-liquid phase separation. In order to investigate the susceptibility of phase separation of nuclear-waste-glasses, two KBS model glasses, ABS-39 and ABS-41, were investigated. Two of the subsequent reports are concerned with this problem. The third report also takes into consideration the effects of MoO 3 on the immiscibility gap. The maximum amount of MoO 3 that can be dissolved in ABS-39 and ABS 41 is also determined. (Auth.)

Basaltic glass alteration in confined media: analogy with nuclear glass in geological disposal conditions

International Nuclear Information System (INIS)

Parruzot, Benjamin

2014-01-01

This dissertation concerns basaltic glass alteration mechanisms and rates. Through a better understanding of the processes controlling the basaltic glass durability, this thesis attempts to establish a link between laboratory studies and volcanic glass alteration in natural environment. The methodology used here is similar to the one used for nuclear glasses. Thus, we measured for the first time the residual alteration rate of basaltic glasses. Protective effect of the alteration film is clearly established. Moreover, synthetic glass representativeness is evaluated through a study focused on the effect of iron oxidation degree on the glass structure and leaching properties. A minor effect of Fe II on the forward rate and a negligible effect on the residual rate are shown. The residual rate is extrapolated at 5 C and compared to the mean alteration rate of natural samples of ages ranging from 1900 to 10 7 years. Non-zeolitized natural glasses follow this linear tendency, suggesting a control of the long-term rate by clayey secondary phase precipitation. Natural environments are open environments: a parametric study was performed in order to quantify the water flow rate effect on chemical composition of the alteration layer. When applied to two natural samples, the obtained laws provide coherent results. It seems possible to unify the descriptive approach from the study of natural environments to the mechanistic approach developed at the laboratory. The next step will consist in developing a model to transpose these results to nuclear glasses. (author) [fr

Hydrogen speciation in hydrated layers on nuclear waste glass

International Nuclear Information System (INIS)

Aines, R.D.; Weed, H.C.; Bates, J.K.

1987-01-01

The hydration of an outer layer on nuclear waste glasses is known to occur during leaching, but the actual speciation of hydrogen (as water or hydroxyl groups) in these layers has not been determined. As part of the Nevada Nuclear Waste Storage Investigations Project, we have used infrared spectroscopy to determine hydrogen speciations in three nuclear waste glass compositions (SRL-131 and 165, and PNL 76-68), which were leached at 90 0 C (all glasses) or hydrated in a vapor-saturated atmosphere at 202 0 C (SRL-131 only). Hydroxyl groups were found in the surface layers of all the glasses. Molecular water was found in the surface of SRL-131 and PNL 76-68 glasses that had been leached for several months in deionized water, and in the vapor-hydrated sample. The water/hydroxyl ratio increases with increasing reaction time; molecular water makes up most of the hydrogen in the thick reaction layers on vapor-phase hydrated glass while only hydroxyl occurs in the least reacted samples. Using the known molar absorptivities of water and hydroxyl in silica-rich glass the vapor-phase layer contained 4.8 moles/liter of molecular water, and 0.6 moles water in the form hydroxyl. A 15 μm layer on SRL-131 glass formed by leaching at 90 0 C contained a total of 4.9 moles/liter of water, 2/3 of which was as hydroxyl. The unreacted bulk glass contains about 0.018 moles/liter water, all as hydroxyl. The amount of hydrogen added to the SRL-131 glass was about 70% of the original Na + Li content, not the 300% that would result from alkali=hydronium ion interdiffusion. If all the hydrogen is then assumed to be added as the result of alkali-H + interdiffusion, the molecular water observed may have formed from condensation of the original hydroxyl groups

Modelling aqueous corrosion of nuclear waste phosphate glass

Energy Technology Data Exchange (ETDEWEB)

Poluektov, Pavel P.; Schmidt, Olga V.; Kascheev, Vladimir A. [Bochvar All-Russian Scientific Research Institute for Inorganic Materials (VNIINM), Moscow (Russian Federation); Ojovan, Michael I., E-mail: m.ojovan@sheffield.ac.uk [Immobilisation Science Laboratory, Department of Materials Science and Engineering, University of Sheffield, Mappin Street, Sheffield, S1 3JD (United Kingdom)

2017-02-15

A model is presented on nuclear sodium alumina phosphate (NAP) glass aqueous corrosion accounting for dissolution of radioactive glass and formation of corrosion products surface layer on the glass contacting ground water of a disposal environment. Modelling is used to process available experimental data demonstrating the generic inhibiting role of corrosion products on the NAP glass surface. - Highlights: • The radionuclides yield is determined by the transport from the glass through the surface corrosion layer. • Formation of the surface layer is due to the dissolution of the glass network and the formation of insoluble compounds. • The model proposed accounts for glass dissolution, formation of corrosion layer, specie diffusion and chemical reactions. • Analytical solutions are found for corrosion layer growth rate and glass components component leaching rates.

Nuclear spin dominated relaxation of atomic tunneling systems in glasses

Energy Technology Data Exchange (ETDEWEB)

Luck, Annina

2016-11-16

The measurements performed in this thesis have revealed a non phononic relaxation channel for atomic tunneling systems in glasses at very low temperatures due to the presence of nuclear electric quadrupoles. Dielectric measurements on the multicomponent glasses N-KZFS11 and HY-1, containing {sup 181}Ta and {sup 165}Ho, respectively, that both carry very large nuclear electric quadrupole moments, show a relaxation rate in the kilohertz range, that is constant for temperatures exceeding the nuclear quadrupole splitting of the relevant isotopes. The results are compared to measurements performed on the glasses Herasil and N-BK7 that both contain no large nuclear quadrupole moments. Using three different setups to measure the complex dielectric function, the measurements cover almost eight orders of magnitude in frequency from 60 Hz to 1 GHz and temperatures down to 7.5 mK. This has allowed us a detailed study of the novel effects observed within this thesis and has led to a simplified model explaining the effects of nuclear electric quadrupoles on the behavior of glasses at low temperatures. Numeric calculations based on this model are compared to the measured data.

Iron Phosphate Glasses: An Alternative for Vitrifying Certain Nuclear Wastes

Energy Technology Data Exchange (ETDEWEB)

Delbert E. Day; Chandra S. Ray; Cheol-Woon Kim

2004-12-28

Vitrification of nuclear waste in a glass is currently the preferred process for waste disposal. DOE currently approves only borosilicate (BS) type glasses for such purposes. However, many nuclear wastes, presently awaiting disposal, have complex and diverse chemical compositions, and often contain components that are poorly soluble or chemically incompatible in BS glasses. Such problematic wastes can be pre-processed and/or diluted to compensate for their incompatibility with a BS glass matrix, but both of these solutions increases the wasteform volume and the overall cost for vitrification. Direct vitrification using alternative glasses that utilize the major components already present in the waste is preferable, since it avoids pre-treating or diluting the waste, and, thus, minimizes the wasteform volume and overall cost.

Iron Phosphate Glasses: An Alternative for Vitrifying Certain Nuclear Wastes

International Nuclear Information System (INIS)

Day, Delbert E.; Ray, Chandra S.; Cheol-Woon Kim

2004-01-01

Vitrification of nuclear waste in a glass is currently the preferred process for waste disposal. DOE currently approves only borosilicate (BS) type glasses for such purposes. However, many nuclear wastes, presently awaiting disposal, have complex and diverse chemical compositions, and often contain components that are poorly soluble or chemically incompatible in BS glasses. Such problematic wastes can be pre-processed and/or diluted to compensate for their incompatibility with a BS glass matrix, but both of these solutions increases the wasteform volume and the overall cost for vitrification. Direct vitrification using alternative glasses that utilize the major components already present in the waste is preferable, since it avoids pre-treating or diluting the waste, and, thus, minimizes the wasteform volume and overall cost

Chemical durability of simulated nuclear glasses containing water

International Nuclear Information System (INIS)

Li, H.; Tomozawa, M.

1995-04-01

The chemical durability of simulated nuclear waste glasses having different water contents was studied. Results from the product consistency test (PCT) showed that glass dissolution increased with water content in the glass. This trend was not observed during MCC-1 testing. This difference was attributed to the differences in reactions between glass and water. In the PCT, the glass network dissolution controlled the elemental releases, and water in the glass accelerated the reaction rate. On the other hand, alkali ion exchange with hydronium played an important role in the MCC-1. For the latter, the amount of water introduced into a leached layer from ion-exchange was found to be much greater than that of initially incorporated water in the glass. Hence, the initial water content has no effect on glass dissolution as measured by the MCC-1 test

Lead iron phosphate glass as a containment medium for disposal of high-level nuclear waste

International Nuclear Information System (INIS)

Boatner, L.A.; Sales, B.C.

1989-01-01

This patent describes lead-iron phosphate glasses containing a high level of Fe 2 O 3 for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste, a highly corrosion resistant, homogeneous, easily processed glass can be formed. For corroding solutions at 90 0 C, with solution pH values in the range between 5 and 9, the corrosion rate of the lead-iron phosphate nuclear waste glass is at least 10 2 to 10 3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. The presence of Fe 2 O 3 in forming the lead-iron phosphate glass is critical. The lead-iron phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear wasteforms

Sulphate solubility and sulphate diffusion in oxide glasses: implications for the containment of sulphate-bearing nuclear wastes; Solubilite et cinetiques de diffusion des sulfates dans differents verres d'oxydes: application au conditionnement des dechets nucleaires sulfates

Energy Technology Data Exchange (ETDEWEB)

Lenoir, M.

2009-09-15

The thesis deals with sulphate solubility and sulphate diffusion in oxide glasses, in order to control sulphate incorporation and sulphate volatilization in nuclear waste glasses. It was conducted on simplified compositions, in the SiO{sub 2}-B{sub 2}O{sub 3}-R{sub 2}O (R = Li, Na, K, Cs), SiO{sub 2}-B{sub 2}O{sub 3}-BaO and V{sub 2}O{sub 5}-B{sub 2}O{sub 3}-BaO systems. These compositions allowed us to study the influence of the nature of network-modifying ions (Li{sup +}, Na{sup +}, K{sup +}, Cs{sup +} or Ba{sup 2+}) and also of former elements (Si, B, V), on structure and properties of glasses. Sulphate volatility is studied in sodium borosilicate melts using an innovative technique of sulphate quantitation with Raman spectroscopy. This technique is useful to obtain kinetic curves of sulphate volatilization. The establishment of a model to fit these curves leads to the determination of diffusion coefficients of sulphate. These diffusion coefficients can thus be compared to diffusion coefficients of other species, determined by other techniques and presented in the literature. They are also linked to diffusion coefficients in relation with the viscosity of the melts. Concerning sulphate solubility in glasses, it depends on glass composition and on the nature of sulphate incorporated. Sulphate incorporation in alkali borosilicate glasses leads to the formation of a sulphate layer floating on top of the melt. Sulphate incorporation in barium borosilicate and boro-vanadate glasses leads to the crystallization of sulphate species inside the vitreous matrix. Moreover, sulphate solubility is higher in these glasses than in alkali borosilicates. Finally, exchanges between cations present in glasses and cations present in the sulphate phase are also studied. (author)

Molecular glasses for nuclear waste encapsulation

International Nuclear Information System (INIS)

Ropp, R.C.

1982-01-01

The use of a molecular glass based upon a polymerized phosphate of aluminum (PAP), indium or gallium overcomes all of the prior objections to use of glass as a high-level nuclear waste (HLW) encapsulation agent. This HLW glass product could not be made to devitrify, dissolved all of the oxides found in calcine, including the difficultly soluble ones, did not form microcrystallites in the melt or subsequent glass-casting, and possessed a hydrolytic etching rate to boiling water even lower than that of HLW-ZBS glass. A precursor compound, M(H 2 PO 4 ) 3 , is prepared, where M is a trivalent metal selected from the group consisting of aluminum, indium and gallium. The impurity level is carefully controlled so as not to exceed 300 ppm total. The precursor crystals may be washed to remove excess phosphoric acid as desired. HLW is added to the crystals and the mixture is then heated at a controlled heating rate to induce solid state polymerization and to form a melt at 1350 degrees C in which the HLW oxides dissolve rapidly

Cold-crucible fabrication of nuclear glasses

International Nuclear Information System (INIS)

Boen, R.

2010-01-01

Vitrification has stood the nuclear industry in good stead, for many years now, as a safe long-term conditioning technology for high-level waste. Major advances are nonetheless still being made, with the development of the cold-crucible technology, affording as it does new possibilities, in terms of volume reduction, and of extending the range of waste products amenable to incorporation. Indeed, by allowing higher melting temperatures to be achieved (1200 - 1400 C degrees), this process opens the way to a considerable increase in glass production capacities, and the fabrication of novel matrices, involving higher incorporation rates than current glasses. In the cold-crucible technology, materials put into the crucible are heated directly through induction. The crucible made of metal is cooled by water circulation. Where the glass comes into contact with the cold wall, a thin layer of solidified glass forms, with a thickness of 5-10 mm preventing the metal forming the crucible from coming into contact with the molten glass. A full scale pilot of the cold crucible was constructed at the La Hague vitrification workshop

Specialty glass development for radiation shielding windows and nuclear waste immobilization

International Nuclear Information System (INIS)

Mandal, S.; Ghorui, S.; Roy Chowdhury, A.; Sen, R.; Chakraborty, A.K.; Sen, S.; Maiti, H.S.

2015-01-01

The technology of two important varieties of specialty glasses, namely high density Radiation Shielding Window (RSW) glass and specialty glass beads of borosilicate composition have been successfully developed in CGCRI with an aim to meet the countries requirement. Radiation Shielding Windows used in nuclear installations, are viewing devices, which allow direct viewing into radioactive areas while still providing adequate protection to the operating personnel. The glass blocks are stabilized against damage from radiation by introducing cerium in definite proportions. Considering the essentially of developing an indigenous technology to make the country self-sufficient for this critical item, CGCRI has taken up a major programme to develop high lead containing glasses required for RSWs under a MoD with BARC. On the other hand, the specialty glass bead of specific composition and properties is a critical material required for management of radioactive waste in a closed nuclear fuel cycle that is followed by India. During reprocessing of the spent nuclear fuel, high level radio-active liquid waste (HLW) is produced containing unwanted radio isotopes some of which remain radioactive for thousands of years. The need is to immobilize them within a molecular structure so that they will not come out and be released to the ambience and thereby needs to be resolved if nuclear power is to make a significant contribution to the country's power requirement. Borosilicate glass has emerged as the material of choice for immobilization due to its unique random network structure

Spectroscopic studies of irradiated glasses: Application in nuclear dosimetry

International Nuclear Information System (INIS)

Farah, Khaled

2010-01-01

The present work aims to study the effects of ionizing radiation on silicate glasses in order to develop a new dosimetry system simple, precise, stable and inexpensive. Indeed, changes in mechanical properties, optical and paramagnetic glasses when subjected to ionizing radiation. The prediction of long-term behavior, physical aging under irradiation, the glass is paramount. many studies have brought many ways to avoid obscuring glass windows used in nuclear reactors or hot cells and optical devices. Recently, much work has concentrated on the application of the color induced by irradiation for developing a recyclable glass in the glass industry is of great interest economically and environmentally.

Leaching of actinides from simulated nuclear waste glass

International Nuclear Information System (INIS)

Pickering, S.; Walker, C.T.; Offermann, P.

1982-01-01

Two types of simulated nuclear waste glass doped with actinides were leached at 200 0 C in distilled water and salt solutions. Am, Np, Pu and U were all preferentially retained in the surface layer on the glass. Leaching ratios of 0.1 to 0.2 for Np and approx. 0.02 for Am were measured. The losses of Am and Np to the leachant were proportional to the total weight loss of the glass and were larger at 10 ml leachant/cm 2 glass than at 5 ml/cm 2 . Weight loss from the glass occurred only at the start of the experiments for periods ranging from 10 h to 10 days according to leachant composition and volume. Wt losses from the C31-3-EC glass were much greater in saturated NaCl solution than in distilled water. Enrichment in the outer surface layer of Al or Ca according to glass type could be correlated with leachant pH, glass composition and weight loss measurements

Fabrication and Characterization of Surrogate Glasses Aimed to Validate Nuclear Forensic Techniques

Science.gov (United States)

2017-12-01

the glass formed during a nuclear event, trinitite [14]. The SiO2 composition is generally greater than 50% for trinitite and can vary appreciably...CHARACTERIZATION OF SURROGATE GLASSES AIMED TO VALIDATE NUCLEAR FORENSIC TECHNIQUES by Ken G. Foos December 2017 Thesis Advisor: Claudia...December 2017 3. REPORT TYPE AND DATES COVERED Master’s thesis 4. TITLE AND SUBTITLE FABRICATION AND CHARACTERIZATION OF SURROGATE GLASSES AIMED TO

Nuclear waste glass melter design including the power and control systems

International Nuclear Information System (INIS)

Chapman, C.C.

1982-01-01

An energy balance of a joule-heated nuclear waste glass melter is used to discuss the problems in the design of the melter geometry and in the specifications of the power and control systems. The relationships between geometry, electrode current density, production rate, load voltage, and load power are presented graphically. The influence of liquid feeding on the surface of the glass and the variability of nuclear waste glass on the design and control during operation is discussed. 10 refs

Lead iron phosphate glass as a containment medium for disposal of high-level nuclear waste

Science.gov (United States)

Boatner, Lynn A.; Sales, Brian C.

1989-01-01

Lead-iron phosphate glasses containing a high level of Fe.sub.2 O.sub.3 for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste, a highly corrosion resistant, homogeneous, easily processed glass can be formed. For corroding solutions at 90.degree. C., with solution pH values in the range between 5 and 9, the corrosion rate of the lead-iron phosphate nuclear waste glass is at least 10.sup.2 to 10.sup.3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. The presence of Fe.sub.2 O.sub.3 in forming the lead-iron phosphate glass is critical. Lead-iron phosphate nuclear waste glass can be prepared at temperatures as low as 800.degree. C., since they exhibit very low melt viscosities in the 800.degree. to 1050.degree. C. temperature range. These waste-loaded glasses do not readily devitrify at temperatures as high as 550.degree. C. and are not adversely affected by large doses of gamma radiation in H.sub.2 O at 135.degree. C. The lead-iron phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear wasteforms.

Gamma radiation induced changes in nuclear waste glass containing Eu

Science.gov (United States)

Mohapatra, M.; Kadam, R. M.; Mishra, R. K.; Kaushik, C. P.; Tomar, B. S.; Godbole, S. V.

2011-10-01

Gamma radiation induced changes were investigated in sodium-barium borosilicate glasses containing Eu. The glass composition was similar to that of nuclear waste glasses used for vitrifying Trombay research reactor nuclear waste at Bhabha Atomic Research Centre, India. Photoluminescence (PL) and electron paramagnetic resonance (EPR) techniques were used to study the speciation of the rare earth (RE) ion in the matrix before and after gamma irradiation. Judd-Ofelt ( J- O) analyses of the emission spectra were done before and after irradiation. The spin counting technique was employed to quantify the number of defect centres formed in the glass at the highest gamma dose studied. PL data suggested the stabilisation of the trivalent RE ion in the borosilicate glass matrix both before and after irradiation. It was also observed that, the RE ion distributes itself in two different environments in the irradiated glass. From the EPR data it was observed that, boron oxygen hole centre based radicals are the predominant defect centres produced in the glass after irradiation along with small amount of E’ centres. From the spin counting studies the concentration of defect centres in the glass was calculated to be 350 ppm at 900 kGy. This indicated the fact that bulk of the glass remained unaffected after gamma irradiation up to 900 kGy.

Composition - structure - properties relationships of peraluminous glasses for nuclear waste containment

International Nuclear Information System (INIS)

Piovesan, Victor

2016-01-01

Part of the Research and Development program concerning high level nuclear waste conditioning aims to assess new glass formulations able to incorporate a high waste content with enhanced properties in terms of homogeneity, thermal stability, long term behavior and process ability. This study focuses on peraluminous glasses, defined by an excess of aluminum ions Al"3"+ in comparison with modifier elements such as Na"+, Li"+ or Ca"2"+. A Design of Experiment approach has been employed to determine relationships between composition of simplified peraluminous glasses (SiO_2 - B_2O_3 - Al_2O_3 - Na_2O - Li_2O - CaO - La_2O_3) and their physical properties such as viscosity, glass transition temperature and glass homogeneity. Moreover, some structural investigation (NMR) was performed in order to better understand the structural role of Na"+, Li"+ and Ca"2"+ and the structural organization of peraluminous glasses. Then, physical and chemical properties of fully simulated peraluminous glasses were characterized to evaluate transposition between simplified and fully simulated glasses and also to put forward the potential of peraluminous glasses for nuclear waste containment. (author) [fr

Solubility of actinides and surrogates in nuclear glasses

International Nuclear Information System (INIS)

Lopez, Ch.

2003-01-01

The nuclear wastes are currently incorporated in borosilicate glass matrices. The resulting glass must be perfectly homogeneous. The work discussed here is a study of actinide (thorium and plutonium) solubility in borosilicate glass, undertaken to assess the extent of actinide solubility in the glass and to understand the mechanisms controlling actinide solubilization. Glass specimens containing; actinide surrogates were used to prepare and optimize the fabrication of radioactive glass samples. These preliminary studies revealed that actinide Surrogates solubility in the glass was enhanced by controlling the processing temperature, the dissolution kinetic of the surrogate precursors, the glass composition and the oxidizing versus reducing conditions. The actinide solubility was investigated in the borosilicate glass. The evolution of thorium solubility in borosilicate glass was determined for temperatures ranging from 1200 deg C to 1400 deg C.Borosilicate glass specimens containing plutonium were fabricated. The experimental result showed that the plutonium solubility limit ranged from 1 to 2.5 wt% PuO 2 at 1200 deg C. A structural approach based on the determination of the local structure around actinides and their surrogates by EXAFS spectroscopy was used to determine their structural role in the glass and the nature of their bonding with the vitreous network. This approach revealed a correlation between the length of these bonds and the solubility of the actinides and their surrogates. (author)

Effect of lead species on the durability of simulated nuclear waste glass

International Nuclear Information System (INIS)

Kuchinski, F.A.

1987-01-01

It has been shown that the incorporation of lead metal into the corrosion environment reduces the leaching rate of nuclear waste glasses. The present study evaluated the effects of lead metal, oxides, alloys, glasses and soluble species on the corrosion rate of a waste glass. The inherent durability of nuclear waste glasses comes from the about due to the insoluble surface film developed during corrosion. This surface film, enriched with iron, aluminum and calcium acts as a diffusion barrier to further corrosion. Except for PbO 2 , all lead species inhibited glass corrosion due to the formation of a surface film enriched in lead. No corroded glass layer was observed below the lead surface layer. Also, no glass corrosion products were found on the lead surface, except for small amounts of silicon. The transport and deposition of lead on the glass surface appears to be the key factors in preventing glass corrosion. At high glass surface area to volume ratios, the glass corroded considerably at short times since the dissolved lead source could not coat the entire glass surface rapidly enough to prevent continued corrosion. Also, experimental solution values did not agree with thermodynamics model predictions. This suggests that kinetic factors, namely diffusion barriers, are controlling the glass corrosion rate

Lead-iron phosphate glass: a stable storage medium for high-level nuclear waste

International Nuclear Information System (INIS)

Sales, B.C.; Boatner, L.A.

1984-01-01

Results are presented which show that lead-iron phosphate glasses are a promising new waste form for the safe immobilization of both high-level defense and high-level commercial radioactive waste. Relative to the borosilicate nuclear waste glasses that are currently the ''reference'' waste form for the long-term disposal of nuclear waste, lead-iron phosphate glasses have several distinct advantages: (1) an aqueous corrosion rate that is about 1000 times lower, (2) a processing temperature that is 100 0 to 250 0 C lower and, (3) a much lower melt viscosity in the temperature range from 800 0 to 1000 0 C. Most significantly, the lead-iron phosphate waste form can be processed using a technology similar to that developed for borosilicate nuclear waste glasses

The role of natural glasses as analogues in projecting the long-term alteration of high-level nuclear waste glasses: Part 1

International Nuclear Information System (INIS)

Mazer, J.J.

1993-01-01

The common observation of glasses persisting in natural environments for long periods of time (up to tens of millions of years) provides compelling evidence that these materials can be kinetically stable in a variety of subsurface environments. This paper reviews how natural and historical synthesized glasses can be employed as natural analogues for understanding and projecting the long-term alteration of high-level nuclear waste glasses. The corrosion of basaltic glass results in many of the same alteration features found in laboratory testing of the corrosion of high-level radioactive waste glasses. Evidence has also been found indicating similarities in the rate controlling processes, such as the effects of silica concentration on corrosion in groundwater and in laboratory leachates. Naturally altered rhyolitic glasses and tektites provide additional evidence that can be used to constrain estimates of long-term waste glass alteration. When reacted under conditions where water is plentiful, the corrosion for these glasses is dominated by network hydrolysis, while the corrosion is dominated by molecular water diffusion and secondary mineral formation under conditions where water contact is intermittent or where water is relatively scarce. Synthesized glasses that have been naturally altered result in alkali-depleted alteration features that are similar to those found for natural glasses and for nuclear waste glasses. The characteristics of these alteration features appear to be dependent on the alteration conditions which affect the dominant reaction processes during weathering. In all cases, care must be taken to ensure that the information being provided by natural analogues is related to nuclear waste glass corrosion in a clear and meaningful way

Stained glasses under the nuclear microprobe: A window into history

Energy Technology Data Exchange (ETDEWEB)

Vilarigues, M. [Dep. de Conservacao e Restauro and R and D Unit Vidro e da Ceramica Para as Artes, FCT-UNL, Quinta da Torre, 2829-516 Caparica (Portugal)], E-mail: mgv@fct.unl.pt; Fernandes, P. [Dep. de Conservacao e Restauro and R and D Unit Vidro e da Ceramica Para as Artes, FCT-UNL, Quinta da Torre, 2829-516 Caparica (Portugal); Alves, L.C.; Silva, R.C. da [Dep. Fisica, LFI, ITN, E.N.10, 2686-953 Sacavem (Portugal)

2009-06-15

Stained glass fragments from the 15th, 16th and 20th centuries, belonging to Mosteiro de Santa Maria da Vitoria, Batalha (Portugal), were characterised non-destructively in a nuclear microprobe. The work aimed at finding the composition of the glasses and glass paintings and relating these with the corresponding production periods. The elemental compositions of the glass fragments were obtained by means of scanning micro-beam Particle Induced X-ray Emission ({mu}-PIXE) spectrometry in selected cross-sections. These were complemented by micro X-Ray fluorescence spectrometry. Characterisation of colour was performed by optical absorption spectroscopy in the UV-vis range, while the corrosion products were identified by optical microscopy and {mu}-FTIR (Fourier Transform Infra Red) spectroscopy in combination with the data generated by {mu}-PIXE. Nuclear microprobe analysis allowed unveiling the compositions and structures, in particular of glass paintings and corrosion products. While it is not surprising that Fe, Cu and Pb were the main elements identified in the grisaille paintings of all studied periods, as well as Ag and Cu found in the glasses decorated with yellow silver painting, their distribution gave important clues on the materials and techniques used to manufacture these stained glasses. Furthermore, it allowed establishing a definite relation between the compositions found and the periods of production, with the added bonus of correctly reassigning the manufacturing period of some samples.

Comprehensive data base of high-level nuclear waste glasses: September 1987 status report: Volume 1, Discussion and glass durability data

International Nuclear Information System (INIS)

Kindle, C.H.; Kreiter, M.R.

1987-12-01

The Materials Characterization Center (MCC) at Pacific Northwest Laboratory is assembling a comprehensive data base (CDB) of experimental data collected for high-level nuclear waste package components. Data collected throughout the world are included in the data base; current emphasis is on waste glasses and their properties. The goal is to provide a data base of properties and compositions and an analysis of dominant property trends as a function of composition. This data base is a resource that nuclear waste producers, disposers, and regulators can use to compare properties of a particular high-level nuclear waste glass product with the properties of other glasses of similar compositions. Researchers may use the data base to guide experimental tests to fill gaps in the available knowledge or to refine empirical models. The data are incorporated into a computerized data base that will allow the data to be extracted based on, for example, glass composition or test duration. 3 figs

Glass produced by underground nuclear explosions

International Nuclear Information System (INIS)

Schwartz, L.; Piwinskii, A.; Ryerson, F.; Tewes, H.; Beiriger, W.

1983-01-01

Detonation of an underground nuclear explosive produces a strong shock wave which propagates spherically outward, vaporizing the explosive and nearby rock and melting, the surrounding rock. The vaporized material expands adiabatically, forming a cavity. As the energy is dissipated during the cavity formation process, the explosive and rock debris condense and mix with the melted rock. The melt flows to the bottom of the cavity where it is quenched by fractured rock fragments falling from above as the cavity collapses. Measurements indicate that about 740 tonnes of rock and/or soil are melted for every kiloton (10 12 calories) of explosive energy, or about 25% of the explosive energy goes to melting rock. The resulting glass composition reflects the composition of the unaltered rock with explosive debris. The appearance ranges from white pumice to dense, dark lava. The bulk composition and color vary with the amount of explosive iron incorporated into the glass. The refractory explosion products are mixed with the solidified melt, although the degree of mixing is variable. Electron microprobe studies of glasses produced by Rainier in welded tuff have produced the following results: glasses are dehydrated relative to the host media, glasses are extremely heterogeneous on a 20 μm scale, a ubiquitous feature is the presence of dark marble-cake regions in the glass, which were locally enriched in iron and may be related to the debris, optically amorphous regions provide evidence of shock melting, only limited major element redistribution and homogenization occur within the cavity

Morphologies of CaMoO sub 4 crystals in simulated nuclear waste disposal glass

Energy Technology Data Exchange (ETDEWEB)

Li, D.; Sengers, E.G.F.; Janssen, F.J.J.G. (KEMA, Arnhem (Netherlands). Chemical Technology and Material Research Dept.); Waal, H. de (TPO-TNO Glass Technology, Eindhoven (Netherlands))

1992-07-01

Fission products can diffuse through nuclear waste disposal glass due to the action of the temperature gradient caused by radioactive decay and the small thermal conductivity of the glass. Diffusion may eventually lead to crystallization. Because the densities of the products of glass crystallization may differ from that of the parent glass, crystallization causes stresses to develop, which can lead to fracture and exposure of increased surface area to environmental attack. Several kinds of crystals including, CaMoO{sub 4}, in the simulated nuclear waste disposal glass K{sub 3}, which consists of Na{sub 2}O, B{sub 2}O{sub 3}, SiO{sub 2} and other oxides, were identified previously after heat treatment. Recently it was found that CaMoO{sub 4} crystals have two kinds of morphologies in glass K{sub 3} heat-treated at temperatures between 870 and 1120 K. One kind of morphology, which is rather special, has not previously been reported. In this letter the morphologies of CaMoO{sub 4} crystals in this simulated nuclear waste disposal glass are discussed. (author).

Influence of nuclear glasses composition on their liability to deterioration

International Nuclear Information System (INIS)

Tovena, I.

1995-01-01

The aim of this thesis is to contribute to the study of the nuclear glasses composition influence on their liability to deterioration. The methodology of the experimental research used has lead to define between the thirty oxides which form the reference glass light water, six oxides of interest. For each of these oxides, a composition variation area has been defined. A matrix of twenty glass compositions has then been defined. The preparation of materials of these compositions has sometimes lead to materials weakly heterogeneous which have been characterized before deterioration. This study has been completed by those of three glasses in a composition variation area narrower of the light water nuclear glass : the R7T7 and two glasses at limits having respectively an initial dissolution velocity at 100 degrees Celsius theoretically maximum and minimum. Some deterioration parameters in pure water have been experimentally measured on the twenty three glasses : 1) an initial dissolution velocity at 100 degrees (Vo 1 00) Celsius and another one at 90 degrees Celsius (Vo 9 0) 2) a dissolution velocity in conditions near the saturation at 90 degrees Celsius 3) an apparent solubility of glass based on the ortho silicic acid activity 4) the evolution of the dissolution kinetics at 90 degrees Celsius in sub-saturated medium towards saturated medium 5) the alteration films nature developed at the glasses surface during these last alteration tests. Some thermodynamic and structural models have been studied in order to predict Vo 9 0 and Vo 1 00. The dissolution kinetic law developed from reference glass dissolution results has been studied with the calculation code LIXIVER. It has not been able to be used for most of the glasses compositions studied. As a consequence, the glasses dissolution control by a surface reaction which are itself controlled by the only dissolved silica is an hypothesis which is not verified for the greater part of the glasses. (O.L.). refs., figs

Fracture toughness in nuclear waste glasses and ceramics: environmental and radiation effects

International Nuclear Information System (INIS)

Weber, W.J.; Matzke, H.J.

1986-03-01

The effects of atmospheric moisture and radiation damage on fracture properties of nuclear waste glasses and ceramics was investigated by indentation techniques. In nuclear waste glasses, atmospheric moisture has no measurable effect on hardness but decreases the fracture toughness; radiation damage, on the other hand, decreased the hardness and increased the fracture toughness. In nuclear ceramics, self-radiation damage from alpha decay decreased the hardness and elastic modules; the fracture toughness increased with dose to a broad maximum and then decreased slightly with further increases in dose

Feasibility Study for Preparation and Use of Glass Grains as an Alternative to Glass Nodules for Vitrification of Nuclear Waste

Energy Technology Data Exchange (ETDEWEB)

Sonavane, M S; Mishra, P.K., E-mail: maheshss@barc.gov.in [Nuclear Recycle Board, Bhabha Atomic Research Centre, Mumbai (India); Mandal, S; Barik, S; Roy Chowdhury, A; Sen, R [Central Glass and Ceramic Institute, Kolkata (India)

2012-10-15

High level nuclear liquid waste (HLW) is immobilized using borosilicate glass matrix. Presently joule heated ceramic melter is being employed for vitrification of HLW in India. Preformed nodules of base glass are fed to melter along with liquid waste in predetermined ratio. In order to reduce the cost incurred for production of glass nodules of base glass, an alternative option of using glass grains was evaluated for its preparation and its suitability for the melter operation. (author)

Feasibility Study for Preparation and Use of Glass Grains as an Alternative to Glass Nodules for Vitrification of Nuclear Waste

International Nuclear Information System (INIS)

Sonavane, M.S.; Mishra, P.K.; Mandal, S.; Barik, S.; Roy Chowdhury, A.; Sen, R.

2012-01-01

High level nuclear liquid waste (HLW) is immobilized using borosilicate glass matrix. Presently joule heated ceramic melter is being employed for vitrification of HLW in India. Preformed nodules of base glass are fed to melter along with liquid waste in predetermined ratio. In order to reduce the cost incurred for production of glass nodules of base glass, an alternative option of using glass grains was evaluated for its preparation and its suitability for the melter operation. (author)

Devitrification of defense nuclear waste glasses: role of melt insolubles

International Nuclear Information System (INIS)

Bickford, D.F.; Jantzen, C.M.

1985-01-01

Time-temperature-transformation (TTT) curves have been determined for simulated nuclear waste glasses bounding the compositional range in the Defense Waste Processing Facility (DWPF). Formulations include all of the minor chemical elements such as ruthenium and chromium which have limited solubility in borosilicate glasses. Heterogeneous nucleation of spinel on ruthenium dioxide, and subsequent nucleation of acmite on spinel is the major devitrification path. Heterogeneous nucleation on melt insolubles causes more rapid growth of crystalline devitrification phases, than in glass free of melt insolubles. These studies point out the importance of simulating waste glass composition and processing as accurately as possible to obtain reliable estimates of glass performance. 11 refs., 8 figs., 1 tab

Nuclear waste under glass, further discussion

Science.gov (United States)

O'Keefe, J. A.; Barkatt, A.; Glass, B. P.; Alterescu, S.

J. J. Crovisier and J. Honnorez [1988] discuss an article by W. W. Maggs, “Mg May Protect Waste Under Glass” [Maggs, 1988] summarizing work by A. Barkatt (Catholic University, Washington, D.C.), B. P. Glass (University of Delaware, Newark), and S. Alterescu and J. A. O'Keefe (NASA/GSFC, Greenbelt, Md.). We found that seawater is orders of magnitude less corrosive t h an fresh water in attacking tektite glass; traced the protective effect to the presence of magnesium, at a level of about 1.3 g/L in seawater; and suggested that the effect might be useful in protecting nuclear waste glasses from corrosion.Crovisier and Honnorez first make the point that the rate of corrosion of glass is, in principle, a function of the ratio of surface area 5 to the effective volume V. This concept, which is usually discussed in American literature under the name of S/V effects, is discussed by Crovisier and Honnorez in terms of the “permeability of the environment.” These effects have been carefully considered throughout our work (see, for example, Barkatt et al. [19867rsqb;). It turns out that in the sea the effective S/V is so small that the effects referred to by Crovisier and Honnorez can be ignored.

Determination of alpha dose rate profile at the HLW nuclear glass/water interface

Energy Technology Data Exchange (ETDEWEB)

Mougnaud, S., E-mail: sarah.mougnaud@cea.fr [CEA Marcoule, DEN/DTCD/SECM, BP 17171, 30207 Bagnols-sur-Cèze cedex (France); Tribet, M.; Rolland, S. [CEA Marcoule, DEN/DTCD/SECM, BP 17171, 30207 Bagnols-sur-Cèze cedex (France); Renault, J.-P. [CEA Saclay, NIMBE UMR 3685 CEA/CNRS, 91191 Gif-sur-Yvette cedex (France); Jégou, C. [CEA Marcoule, DEN/DTCD/SECM, BP 17171, 30207 Bagnols-sur-Cèze cedex (France)

2015-07-15

Highlights: • The nuclear glass/water interface is studied. • The way the energy of alpha particles is deposited is modeled using MCNPX code. • A model giving dose rate profiles at the interface using intrinsic data is proposed. • Bulk dose rate is a majoring estimation in alteration layer and in surrounding water. • Dose rate is high in small cracks; in larger ones irradiated volume is negligible. - Abstract: Alpha irradiation and radiolysis can affect the alteration behavior of High Level Waste (HLW) nuclear glasses. In this study, the way the energy of alpha particles, emitted by a typical HLW glass, is deposited in water at the glass/water interface is investigated, with the aim of better characterizing the dose deposition at the glass/water interface during water-induced leaching mechanisms. A simplified chemical composition was considered for the nuclear glass under study, wherein the dose rate is about 140 Gy/h. The MCNPX calculation code was used to calculate alpha dose rate and alpha particle flux profiles at the glass/water interface in different systems: a single glass grain in water, a glass powder in water and a water-filled ideal crack in a glass package. Dose rate decreases within glass and in water as distance to the center of the grain increases. A general model has been proposed to fit a dose rate profile in water and in glass from values for dose rate in glass bulk, alpha range in water and linear energy transfer considerations. The glass powder simulation showed that there was systematic overlapping of radiation fields for neighboring glass grains, but the water dose rate always remained lower than the bulk value. Finally, for typical ideal cracks in a glass matrix, an overlapping of irradiation fields was observed while the crack aperture was lower than twice the alpha range in water. This led to significant values for the alpha dose rate within the crack volume, as long as the aperture remained lower than 60 μm.

The glass model of Muelheim-Kaerlich nuclear power station

International Nuclear Information System (INIS)

Kuttruf, H.; Lemke, W.

1986-01-01

The glass model represents the nuclear steam generator system of Muelheim-Kaerlich nuclear power station on a scale of 1:25 and in simplified form, so that the thermohydraulic behaviour in both normal operational and fault conditions can be represented. A set-up time of about one hour results in a helpful aid to instruction. (orig.) [de

The role of nuclear analytical techniques in the study of aqueous corrosion of glasses

International Nuclear Information System (INIS)

Trocellier, P.

1984-01-01

Direct observation of resonant nuclear reactions, backscattering spectrometry and X ray microanalysis with a nuclear microprobe were used to determine elementary depth profiles in the near surface region of leached glasses. Some computing programs required to interpretate the analytical information detected were built. Experimental conditions to characterize glass samples without secondary effects were defined; and the influence of some leaching parameters was studied to describe the first stages of aqueous corrosion of borosilicate glasses [fr

Stress corrosion in a borosilicate glass nuclear wasteform

International Nuclear Information System (INIS)

Ringwood, A.E.; Willis, P.

1984-01-01

The authors discuss a typical borosilicate glass wasteform which, when exposed to water vapour and water for limited periods, exhibits evidence of stress corrosion cracking arising from the interaction of polar OH groups with stressed glass surfaces. Glass wasteforms may experience similar stress corrosion cracking when buried in a geological repository and exposed to groundwaters over an extended period. This would increase the effective surface areas available for leaching by groundwater and could decrease the lifetime of the wasteform. Conventional leach-testing methods are insensitive to the longer-term effects of stress corrosion cracking. It is suggested that specific fracture-mechanics tests designed to evaluate susceptibility to stress corrosion cracking should be used when evaluating the wasteforms for high-level nuclear wastes. (author)

Study of archaeological analogs for the validation of nuclear glass long-term behavior models

International Nuclear Information System (INIS)

Verney-Carron, A.

2008-10-01

Fractured archaeological glass blocks collected from a shipwreck discovered in the Mediterranean Sea near Embiez Island (Var) were investigated because of their morphological analogy with vitrified nuclear waste and of a known and stable environment. These glasses are fractured due to a fast cooling after they were melted (like nuclear glass) and have been altered for 1800 years in seawater. This work results in the development and the validation of a geochemical model able to simulate the alteration of a fractured archaeological glass block over 1800 years. The kinetics associated with the different mechanisms (interdiffusion and dissolution) and the thermodynamic parameters of the model were determined by leaching experiments. The model implemented in HYTEC software was used to simulate crack alteration over 1800 years. The consistency between simulated alteration thicknesses and measured data on glass blocks validate the capacity of the model to predict long-term alteration. This model is able to account for the results from the characterization of crack network and its state of alteration. The cracks in the border zone are the most altered due to a fast renewal of the leaching solution, whereas internal cracks are thin because of complex interactions between glass alteration and transport of elements in solution (influence of initial crack aperture and of the crack sealing). The lowest alteration thicknesses, as well as their variability, can be explained. The analog behavior of archaeological and nuclear glasses from leaching experiments makes possible the transposition of the model to nuclear glass in geological repository. (author)

Chemical speciation of U, Fe, and Pu in melt glass from nuclear weapons testing

Energy Technology Data Exchange (ETDEWEB)

Pacold, J. I.; Lukens, W. W.; Booth, C. H.; Shuh, D. K. [Chemical Sciences Division, Lawrence Berkeley National Laboratory, Berkeley, California 94720 (United States); Knight, K. B.; Eppich, G. R. [Nuclear and Chemical Sciences Division, Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Holliday, K. S. [Materials Science Division, Lawrence Livermore National Laboratory, Livermore, California 94550 (United States)

2016-05-21

Nuclear weapons testing generates large volumes of glassy materials that influence the transport of dispersed actinides in the environment and may carry information on the composition of the detonated device. We determine the oxidation state of U and Fe (which is known to buffer the oxidation state of actinide elements and to affect the redox state of groundwater) in samples of melt glass collected from three U.S. nuclear weapons tests. For selected samples, we also determine the coordination geometry of U and Fe, and we report the oxidation state of Pu from one melt glass sample. We find significant variations among the melt glass samples and, in particular, find a clear deviation in one sample from the expected buffering effect of Fe(II)/Fe(III) on the oxidation state of uranium. In the first direct measurement of Pu oxidation state in a nuclear test melt glass, we obtain a result consistent with existing literature that proposes Pu is primarily present as Pu(IV) in post-detonation material. In addition, our measurements imply that highly mobile U(VI) may be produced in significant quantities when melt glass is quenched rapidly following a nuclear detonation, though these products may remain immobile in the vitrified matrices. The observed differences in chemical state among the three samples show that redox conditions can vary dramatically across different nuclear test conditions. The local soil composition, associated device materials, and the rate of quenching are all likely to affect the final redox state of the glass. The resulting variations in glass chemistry are significant for understanding and interpreting debris chemistry and the later environmental mobility of dispersed material.

Chemical speciation of U, Fe, and Pu in melt glass from nuclear weapons testing

Science.gov (United States)

Pacold, J. I.; Lukens, W. W.; Booth, C. H.; Shuh, D. K.; Knight, K. B.; Eppich, G. R.; Holliday, K. S.

2016-05-01

Nuclear weapons testing generates large volumes of glassy materials that influence the transport of dispersed actinides in the environment and may carry information on the composition of the detonated device. We determine the oxidation state of U and Fe (which is known to buffer the oxidation state of actinide elements and to affect the redox state of groundwater) in samples of melt glass collected from three U.S. nuclear weapons tests. For selected samples, we also determine the coordination geometry of U and Fe, and we report the oxidation state of Pu from one melt glass sample. We find significant variations among the melt glass samples and, in particular, find a clear deviation in one sample from the expected buffering effect of Fe(II)/Fe(III) on the oxidation state of uranium. In the first direct measurement of Pu oxidation state in a nuclear test melt glass, we obtain a result consistent with existing literature that proposes Pu is primarily present as Pu(IV) in post-detonation material. In addition, our measurements imply that highly mobile U(VI) may be produced in significant quantities when melt glass is quenched rapidly following a nuclear detonation, though these products may remain immobile in the vitrified matrices. The observed differences in chemical state among the three samples show that redox conditions can vary dramatically across different nuclear test conditions. The local soil composition, associated device materials, and the rate of quenching are all likely to affect the final redox state of the glass. The resulting variations in glass chemistry are significant for understanding and interpreting debris chemistry and the later environmental mobility of dispersed material.

Study of film semiconductor glass-metal interfaces by nuclear methods

International Nuclear Information System (INIS)

Wehr, Muryel.

1979-01-01

The use of nuclear method analysis, particularly α particles and Li + ions elastic backscattering permitted to study the glass chalcogenide-metal interdiffusion submitted to thermal and electric stresses. The 8 MeV alpha particles are of a great interest, they increase five times the depth of the gold analysis in glasses compared with the 3,5 MeV alpha particles [fr

Composition models for the viscosity and chemical durability of West Valley related nuclear waste glasses

International Nuclear Information System (INIS)

Feng, X.; Saad, E.E.; Freeborn, W.P.; Macedo, P.B.; Pegg, I.L.; Sassoon, R.E.; Barkatt, A.; Finger, S.M.

1988-01-01

There are two important criteria that must be satisfied by a nuclear waste glass durability and processability. The chemical composition of the glass must be such that it does not dissolve or erode appreciably faster than the decay of the radioactive materials embedded in it. The second criterion, processability, means that the glass must melt with ease, must be easily pourable, and must not crystallize appreciably. This paper summarizes the development of simple models for predicting the durability and viscosity of nuclear waste glasses from their composition

Redox reaction and foaming in nuclear waste glass melting

Energy Technology Data Exchange (ETDEWEB)

Ryan, J.L.

1995-08-01

This document was prepared by Pacific Northwest Laboratory (PNL) and is an attempt to analyze and estimate the effects of feed composition variables and reducing agent variables on the expected chemistry of reactions occurring in the cold cap and in the glass melt in the nuclear waste glass Slurry-fed, joule-heated melters as they might affect foaming during the glass-making process. Numerous redox reactions of waste glass components and potential feed additives, and the effects of other feed variables on these reactions are reviewed with regard to their potential effect on glass foaming. A major emphasis of this report is to examine the potential positive or negative aspects of adjusting feed with formic acid as opposed to other feed modification techniques including but not limited to use of other reducing agents. Feed modification techniques other than the use of reductants that should influence foaming behavior include control of glass melter feed pH through use of nitric acid. They also include partial replacement of sodium salts by lithium salts. This latter action (b) apparently lowers glass viscosity and raises surface tension. This replacement should decrease foaming by decreasing foam stability.

Development of Models to Predict the Redox State of Nuclear Waste Containment Glass

Energy Technology Data Exchange (ETDEWEB)

Pinet, O.; Guirat, R.; Advocat, T. [Commissariat a l' Energie Atomique (CEA), Departement de Traitement et de Conditionnement des Dechets, Marcoule, BP 71171, 30207 Bagnols-sur-Ceze Cedex (France); Phalippou, J. [Universite de Montpellier II, Laboratoire des Colloides, Verres et Nanomateriaux, 34095 Montpellier Cedex 5 (France)

2008-07-01

Vitrification is one of the recommended immobilization routes for nuclear waste, and is currently implemented at industrial scale in several countries, notably for high-level waste. To optimize nuclear waste vitrification, research is conducted to specify suitable glass formulations and develop more effective processes. This research is based not only on experiments at laboratory or technological scale, but also on computer models. Vitrified nuclear waste often contains several multi-valent species whose oxidation state can impact the properties of the melt and of the final glass; these include iron, cerium, ruthenium, manganese, chromium and nickel. Cea is therefore also developing models to predict the final glass redox state. Given the raw materials and production conditions, the model predicts the oxygen fugacity at equilibrium in the melt. It can also estimate the ratios between the oxidation states of the multi-valent species contained in the molten glass. The oxidizing or reductive nature of the atmosphere above the glass melt is also taken into account. Unlike the models used in the conventional glass industry based on empirical methods with a limited range of application, the models proposed are based on the thermodynamic properties of the redox species contained in the waste vitrification feed stream. The thermodynamic data on which the model is based concern the relationship between the glass redox state and the oxygen fugacity in the molten glass. The model predictions were compared with oxygen fugacity measurements for some fifty glasses. The experiments carried out at laboratory and industrial scale with a cold crucible melter. The oxygen fugacity of the glass samples was measured by electrochemical methods and compared with the predicted value. The differences between the predicted and measured oxygen fugacity values were generally less than 0.5 Log unit. (authors)

Rhyolitic glasses as natural analogues of nuclear waste glasses: behaviour of an Icelandic glass upon natural aqueous corrosion

International Nuclear Information System (INIS)

Magonthier, M.-C.; Petit, J.-C.; Dran, J.-C.

1992-01-01

A detailed study of the altered rims present in narrow fissures of a 52 ka-old Icelandic obsidian reveals the behaviour of transition and heavy elements, as well as the mechanism and kinetics of alteration, during glass/solution interaction. These complex altered rims are alkali depleted and consist of alternating layers of Fe-rich aluminosilicate and aluminium thihydroxide. The elemental partitioning observed on this naturally corroded obsidian is supported by laboratory experiments performed on the same glass, the elemental accumulation being explained by the formation of a hydrosilicate. A good correlation exists between the thickness of the altered rims and that calculated from the amounts of Fe and Ti accumulated locally. Thus, immobile elements can be used reliably as indices of the extent of alteration because only near-equilibrium conditions occur. The good agreement between the experimental hydration rate of obsidians and the progress of natural corrosion, leads to the assumption that ion diffusion is the long-term controlling mechanism of corrosion. Such an assumption is supported by the particular distribution of the immobile elements which is due to ion diffusion and coprecipitation processes (self-organization genesis). These observations have implications for nuclear waste disposal topics and support the validity of obsidians as analogues of nuclear waste glasses with respect to some local environmental constraints induced by waste packaging and disposal. (author)

Nanoporous Glasses for Nuclear Waste Containment

OpenAIRE

Woignier, Thierry; Primera, Juan; Reynes, Jerôme

2016-01-01

Research is in progress to incorporate nuclear waste in new matrices with high structural stability, resistance to thermal shock, and high chemical durability. Interactions with water are important for materials used as a containment matrix for the radio nuclides. It is indispensable to improve their chemical durability to limit the possible release of radioactive chemical species, if the glass structure is attacked by corrosion. By associating high structural stability and high chemical dura...

Nuclear waste glass product consistency test (PCT): Version 7.0. Revision 3

International Nuclear Information System (INIS)

Jantzen, C.M.; Bibler, N.E.; Beam, D.C.; Ramsey, W.G.

1994-06-01

Liquid high-level nuclear waste will be immobilized at the Savannah River Site (SRS) by vitrification in borosilicate glass. The glass will be produced in the Defense Waste Processing Facility (DWPF), poured into stainless steel canisters, and eventually disposed of in a geologic repository. In order to comply with the Waste Acceptance Product Specifications (WAPS), the durability of the glass needs to be measured during production to assure its long term stability and radionuclide release properties. A durability test, designated the Product Consistency Test (PCT), was developed for DWPF glass in order to meet the WAPS requirements. The response of the PCT procedure was based on extensive testing with glasses of widely different compositions. The PCT was determined to be very reproducible, to yield reliable results rapidly, and to be easily performed in shielded cell facilities with radioactive samples. Version 7.0 of the PCT procedure is attached. This draft version has been submitted to ASTM for full committee (C26, Nuclear Fuel Cycle) ballot after being balloted successfully through subcommittee C26.13 on Repository Waste Package Materials Testing

Thermo-chemistry of nuclear waste glasses: a new approach

International Nuclear Information System (INIS)

Linard, Y.; Neuville, D.R.; Richet, P.

1997-01-01

Understanding of the stability and weathering of glasses used for storing fission products is hampered by a general lack of basic thermochemical information. Models have been setup to predict Gibbs free energies of dissolution of glasses, but ascertaining their accuracy is made difficult by the very lack of reliable experimental data with which model results should be compared. As enthalpies of formation can in principle be determined from usual solution calorimetry experiments, the lack of Gibbs-free energy data for glasses mainly stems from the fact that, as disordered substances, glasses do not obey the third principle and have indeed large configurational entropies. These entropies can be determined from thermochemical measurements only when there exist a congruently melting crystalline compound with the same composition. Using available data, we have calculated the Gibbs-free energies of formation of a series of silicate glasses for which such a calorimetric determination is possible. With these results, we assess the predictions of Paul's model (1977) for calculating Gibbs-free energies of dissolution. As the complex compositions of the borosilicate glasses used for nuclear waste storage prevent determining configurational entropies by calorimetric methods, we point out how these can be determined instead from viscosity measurements. We finally discuss the implications of this approach for modeling of water-glass interactions. (authors)

Secondary phases formed during nuclear waste glass-water interactions: Thermodynamic and derived properties

International Nuclear Information System (INIS)

McKenzie, W.F.

1992-08-01

The thermodynamic properties of secondary phases observed to form during nuclear waste glass-water interactions are of particular interest as it is with the application of these properties together with the thermodynamic properties of other solid phases, fluid phases, and aqueous species that one may predict the environmental consequences of introducing radionuclides contained in the glass into groundwater at a high-level nuclear waste repository. The validation of these predicted consequences can be obtained from laboratory experiments and field observations at natural analogue sites. The purpose of this report is to update and expand the previous compilation (McKenzie, 1991) of thermodynamic data retrieved from the literature and/or estimated for secondary phases observed to form (and candidate phases from observed chemical compositions) during nuclear waste glass-water interactions. In addition, this report includes provisionally recommended thermodynamic data of secondary phases

West Valley high-level nuclear waste glass development: a statistically designed mixture study

Energy Technology Data Exchange (ETDEWEB)

Chick, L.A.; Bowen, W.M.; Lokken, R.O.; Wald, J.W.; Bunnell, L.R.; Strachan, D.M.

1984-10-01

The first full-scale conversion of high-level commercial nuclear wastes to glass in the United States will be conducted at West Valley, New York, by West Valley Nuclear Services Company, Inc. (WVNS), for the US Department of Energy. Pacific Northwest Laboratory (PNL) is supporting WVNS in the design of the glass-making process and the chemical formulation of the glass. This report describes the statistically designed study performed by PNL to develop the glass composition recommended for use at West Valley. The recommended glass contains 28 wt% waste, as limited by process requirements. The waste loading and the silica content (45 wt%) are similar to those in previously developed waste glasses; however, the new formulation contains more calcium and less boron. A series of tests verified that the increased calcium results in improved chemical durability and does not adversely affect the other modeled properties. The optimization study assessed the effects of seven oxide components on glass properties. Over 100 melts combining the seven components into a wide variety of statistically chosen compositions were tested. Viscosity, electrical conductivity, thermal expansion, crystallinity, and chemical durability were measured and empirically modeled as a function of the glass composition. The mathematical models were then used to predict the optimum formulation. This glass was tested and adjusted to arrive at the final composition recommended for use at West Valley. 56 references, 49 figures, 18 tables.

Nuclear waste glass product consistency test (PCT), Version 5.0

International Nuclear Information System (INIS)

Jantzen, C.M.; Bibler, N.E.; Beam, D.C.; Ramsey, W.G.; Waters, B.J.

1992-06-01

Liquid high-level nuclear waste will be immobilized at the Savannah River Site (SRS) by vitrification in borosilicate glass. The glass will be produced in the Defense Waste Processing Facility (DWPF), poured into stainless steel canisters, and eventually disposed of in a geologic repository. In order to comply with the Waste Acceptance Preliminary Specifications (WAPS), the durability of the glass needs to be measured during production to assure its long term stability and radionuclide release properties. A durability test, designated the Produce Consistency Test (PCT), was developed for DWPF glass in order to meet the WAPS requirements. The response of the PCT procedure was based on extensive testing with glasses of widely different compositions. The PCT was determined to be very reproducible, to yield reliable results rapidly, and to be easily performed in shielded cell facilities with radioactive samples. Version 5.0 of the PCT procedure is attached

Nano-Continuum Modeling of a Nuclear Glass Specimen Altered for 25 Years

Energy Technology Data Exchange (ETDEWEB)

Steefel, Carl

2014-01-06

The purpose of this contribution is to report on preliminary nano-continuum scale modeling of nuclear waste glass corrosion. The focus of the modeling is an experiment involving a French glass SON68 specimen leached for 25 years in a granitic environment. In this report, we focus on capturing the nano-scale concentration profiles. We use a high resolution continuum model with a constant grid spacing of 1 nanometer to investigate the glass corrosion mechanisms.

Contribution to the study of the effects of α-irradiation in nuclear glasses

International Nuclear Information System (INIS)

Abbas, A.

2001-01-01

The main topic of this work is to characterise the effects of α-disintegration in nuclear waste glasses. Experimental and numerical approaches have been considered. The structure of the French nuclear waste glass (R7T7) has been simulated using four- and six-oxides simplified glasses which contain the main elements of the R7T7 glass: SiO 2 , B 2 O 3 , Na 2 O, ZrO 2 , Al 2 O 3 and CaO. Four- and six-oxides glasses have been irradiated with 1 MeV-He + (ionisation) and 2.1 MeV-Kr 3+ (ionisation and atomic collisions) ions in order to reproduce the effects of the α-particle and of the recoil nucleus emitted during α-disintegration of actinides, and also to differentiate electronic and ballistic effects. Irradiated glasses have been characterised using several techniques, which have been adapted to the peculiarities of our samples (isolated material, small irradiated depth). The results point out the salient role of sodium in the observed modifications: depth concentration profiles obtained with RBS show an accumulation of sodium at the irradiated surface. We found a apparent acceleration of sodium release in leaching experiments which confirm that point. Modifications observed in Raman spectra of irradiated glasses show an increase of the polymerisation (increase of Q 3 /Q 2 ratio) due to sodium migration. In simplified glasses we have found that the modifications of mechanical properties by external irradiations reproduce the modifications observed in actinide doped nuclear glass (decrease of hardness and increase of fracture toughness). At the same time, we performed Molecular Dynamics simulations of a six-oxides glass. We have shown that the surface modifies the glass structure down to a depth of 10 Angstrom: modification of depth concentration profiles, decrease of the atomic coordination number (A1, B and Si). During cascades, we found that atomic displacements are easier near the surface. This behaviour is also observed when the glass is submitted to an

Contributions of vitreous natural analogs to the investigation of long-term nuclear glass behavior

International Nuclear Information System (INIS)

Techer, I.

1999-01-01

This study assesses the extend of the analogy between the alteration behavior in water and in a moist clay environment of aluminosilicate volcanic glass and alumino-borosilicate nuclear containment glass. Basaltic glass alteration in water initially occurs by hydrolysis processes with an activation energy on the order of 73 kJ.mol -1 . As the reaction progresses, the alteration rate drops by over four orders of magnitude from the initial rate r 0 , The alteration kinetics are not governed by the alteration solution chemistry alone, the glass alteration film appears to have a major role as a diffusion barrier limiting the transfer of reaction species and products. All these aspects highlight the behavioral analogy between basaltic glass and nuclear borosilicate glass in aqueous media. Conversely, the alteration reaction of obsidian-type volcanic glass involves other mechanisms than those governing the dissolution of borosilicate glass. Basaltic glass alteration is also examined in the presence of a clay environmental material, in a study of the natural basaltic glass and argillaceous pelites system of the Salagou basin in southern France, in an approach combining mineralogical, chemical and isotopic data to assess the interactions between a basaltic glass and the argillaceous pelites. Laboratory leach test results with basaltic glass and measured data for the Salagou glass in its natural environment are modeled using a code implementing a kinetic law coupling diffusive transfer of dissolved silica with a reaction affinity law. (author)

Design and test of the borosilicate glass burnable poison rod for Qinshan nuclear power plant core

International Nuclear Information System (INIS)

Huang Jinhua; Sun Hanhong

1988-08-01

Material for the burnable poison of Qinshan Nuclear Power Plant core is GG-17 borosilicate glass. The chemical composition and physico-chemical properties of GG-17 is very close to Pyrex-7740 glass used by Westinghouse. It is expected from the results of the experiments that the borosilicate glass burnable poison rod can be successfully used in Qinshan Nuclear Power Plant due to good physical, mechanical, corrosion-resistant and irradiaton properties for both GG-17 glass and cold-worked stainless steel cladding. Change of material for burnable poison from boron-bearing stainless steel to borosilicate glass will bring about much more economic benefit to Qinshan Naclear Power Plant

Calculation of the viscosity of nuclear waste glass systems

International Nuclear Information System (INIS)

Shah, R.; Behrman, E.C.; Oksoy, D.

1990-01-01

Viscosity is one of the most important processing parameters and one of the most difficult to calculate theoretically, particularly for multicomponent systems like nuclear waste glasses. Here, the authors propose a semi-empirical approach based on the Fulcher equation, involving identification of key variables, for which coefficients are then determined by regression analysis. Results are presented for two glass systems, and compared to results of previous workers and to experiment. The authors also sketch a first-order statistical mechanical perturbation theory calculation for the effects on viscosity of a change in composition of the melt

Lead-iron phosphate glass as a containment medium for the disposal of high-level nuclear wastes

Science.gov (United States)

Boatner, L.A.; Sales, B.C.

1984-04-11

Disclosed are lead-iron phosphate glasses containing a high level of Fe/sub 2/O/sub 3/ for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste

Comparative study of seven glasses for solidification of nuclear wastes

International Nuclear Information System (INIS)

Nogues, J.L.; Hench, L.L.; Zarzycki, J.

1982-06-01

The relative leaching behavior of seven alkali borosilicate glasses considered for immobilization of high level radioactive wastes was compared using a static 90 0 C leach test. Leaching times studied were 1, 3, 7, 14 and 28 days with ratios of glass surface area (SA) to solution volume (V) being SA/V = 1.0 cm -1 and 0.1 cm -1 . With the range of glass compositions studied, it was not possible to determine the effect of each element on leaching behavior, however some conclusions regarding the general influence of the glass network formers can be made: the addition of Al 2 O 3 , results in a large increase in the chemical durability of the glass. The presence of Fe 2 O 3 , is necessary to develop with Al 2 O 3 a second protective layer on top of the silica-rich film that results from rapid dealkalization. The difference between the results obtained at SA/V = 1.0 cm -1 and 0.1 cm -1 shows the importance of understanding both the effects of glass composition and solution concentrations on the behavior of nuclear waste glasses

Glass: a candidate engineered material for management of high level nuclear waste

International Nuclear Information System (INIS)

Mishra, R.K.; Kaushik, C.P.

2011-01-01

While the commercial importance of glass is generally recognized, a few people are aware of extremely wide range of glass formulations that can be made and of the versatility of this engineered material. Some of the recent developments in the field of glass leading to various technological applications include glass fiber reinforcement of cement to give new building materials, substrates for microelectronics circuitry in form of semiconducting glasses, nuclear waste immobilization and specific medical applications. The present paper covers fundamental understanding of glass structure and its application for immobilization of high level radioactive liquid waste. High level radioactive liquid waste (HLW) arising during reprocessing of spent fuel are immobilized in sodium borosilicate glass matrix developed indigenously. Glass compositions are modified according to the composition of HLW to meet the criteria of desirable properties in terms. These glass matrices have been characterized for different properties like homogeneity, chemical durability, thermal stability and radiation stability. (author)

Calcium titanium silicate based glass-ceramic for nuclear waste immobilisation

Science.gov (United States)

Sharma, K.; Srivastav, A. P.; Goswami, M.; Krishnan, Madangopal

2018-04-01

Titanate based ceramics (synroc) have been studied for immobilisation of nuclear wastes due to their high radiation and thermal stability. The aim of this study is to synthesis glass-ceramic with stable phases from alumino silicate glass composition and study the loading behavior of actinides in glass-ceramics. The effects of CaO and TiO2 addition on phase evolution and structural properties of alumino silicate based glasses with nominal composition x(10CaO-9TiO2)-y(10Na2O-5 Al2O3-56SiO2-10B2O3); where z = x/y = 1.4-1.8 are reported. The glasses are prepared by melt-quench technique and characterized for thermal and structural properties using DTA and Raman Spectroscopy. Glass transition and peak crystallization temperatures decrease with increase of CaO and TiO2 content, which implies the weakening of glass network and increased tendency of glasses towards crystallization. Sphene (CaTiSiO5) and perovskite (CaTiO3) crystalline phases are confirmed from XRD which are well known stable phase for conditioning of actinides. The microsturcture and elemental analysis indicate the presence of actinide in stable crystalline phases.

Molecular Dynamics-based Simulations of Bulk/Interfacial Structures and Diffusion Behaviors in Nuclear Waste Glasses

Energy Technology Data Exchange (ETDEWEB)

Du, Jincheng

2018-03-16

This NEUP Project aimed to generate accurate atomic structural models of nuclear waste glasses by using large-scale molecular dynamics-based computer simulations and to use these models to investigate self-diffusion behaviors, interfacial structures, and hydrated gel structures formed during dissolution of these glasses. The goal was to obtain realistic and accurate short and medium range structures of these complex oxide glasses, to provide a mechanistic understanding of the dissolution behaviors, and to generate reliable information with predictive power in designing nuclear waste glasses for long-term geological storage. Looking back of the research accomplishments of this project, most of the scientific goals initially proposed have been achieved through intensive research in the three and a half year period of the project. This project has also generated a wealth of scientific data and vibrant discussions with various groups through collaborations within and outside of this project. Throughout the project one book chapter and 14 peer reviewed journal publications have been generated (including one under review) and 16 presentations (including 8 invited talks) have been made to disseminate the results of this project in national and international conference. Furthermore, this project has trained several outstanding graduate students and young researchers for future workforce in nuclear related field, especially on nuclear waste immobilization. One postdoc and four PhD students have been fully or partially supported through the project with intensive training in the field material science and engineering with expertise on glass science and nuclear waste disposal

Glass-Metal Joining in Nuclear Environment: the State of the Art

International Nuclear Information System (INIS)

Jacobs, M.

2007-01-01

Full text of publication follows: In the ITER fusion machine and in material testing fission reactors, it is not possible to avoid the use of non-metallic materials like glass for example. There is therefore a need to apply metal to glass joints. This problem arose already at the beginning of the 19. century when the electric light bulb was invented. Nowadays this type of glass-metal joint is very successful and widely used in the electronic industry. In the case of ITER and material testing reactors, glass-metal joints are necessary for the fixation of the optical windows and optical fibres to a metal structure to perform diagnostics. These types of joints are still difficult to make and their behaviour is not fully understood. A joint between glass and metal for a nuclear or fusion application has indeed to resist high temperatures and high neutron fluences, while keeping a good mechanical strength and remaining leak tight. These characteristics are difficult to obtain under these severe conditions. This paper presents an overview of the different joining technologies that can be used to join glass to metal in a severe nuclear environment. The working mechanism of the technologies are explained, together with their respective advantages and drawbacks. Three different types of joining are discussed: fastening, liquid phase joining and solid phase joining. Fastening is a mechanical attachment technique, not achieving easily hermetic seals. Liquid and solid phase joining on the other hand form a real bond, what makes the joint much stronger. The most important technologies using liquid phase joining are adhesive bonding, fusion welding and brazing. In the case of the solid phase joining the choices are ultrasonic torsion welding, diffusion bonding and electrostatic bonding. If it is usually not possible to join the glass directly to the metal, an interlayer must be used. One speaks then of indirect joining. The paper will conclude with a discussion on the best

Processing glass-pyrochlore composites for nuclear waste encapsulation

International Nuclear Information System (INIS)

Pace, S.; Cannillo, V.; Wu, J.; Boccaccini, D.N.; Seglem, S.; Boccaccini, A.R.

2005-01-01

Glass matrix composites have been developed as alternative materials to immobilize nuclear solid waste, in particular actinides. These composites are made of soda borosilicate glass matrix, into which particles of lanthanum zirconate pyrochlore are encapsulated in concentrations of 30 vol.%. The fabrication process involves powder mixing followed by hot-pressing. At the relatively low processing temperature used (620 deg. C), the pyrochlore crystalline structure of the zirconate, which is relevant for containment of radioactive nuclei, remains unaltered. The microstructure of the composites exhibits a homogeneous distribution of isolated pyrochlore particles in the glass matrix and strong bonding at the matrix-particle interfaces. Hot-pressing was found to lead to high densification (95% th.d.) of the composite. The materials are characterized by relatively high elastic modulus, flexural strength, hardness and fracture toughness. A numerical approach using a microstructure-based finite element solver was used in order to investigate the mechanical properties of the composites

Natural glass analogues to alteration of nuclear waste glass: A review and recommendations for further study

International Nuclear Information System (INIS)

McKenzie, W.F.

1990-01-01

The purpose of this report is to review previous work on the weathering of natural glasses; and to make recommendations for further work with respect to studying the alteration of natural glasses as it relates quantifying rates of dissolution. the first task was greatly simplified by the published papers of Jercinovic and Ewing (1987) and Byers, Jercinovic, and Ewing (1987). The second task is obviously the more difficult of the two and the author makes no claim of completeness in this regard. Glasses weather in the natural environment by reacting with aqueous solutions producing a rind of secondary solid phases. It had been proposed by some workers that the thickness of this rind is a function of the age of the glass and thus could be used to estimate glass dissolution rates. However, Jercinovic and Ewing (1987) point out that in general the rind thickness does not correlate with the age of the glass owing to the differences in time of contact with the solution compared to the actual age of the sample. It should be noted that the rate of glass dissolution is also a function of the composition of both the glass and the solution, and the temperature. Quantification of the effects of these parameters (as well as time of contact with the aqueous phase and flow rates) would thus permit a prediction of the consequences of glass-fluid interactions under varying environmental conditions. Defense high- level nuclear waste (DHLW), consisting primarily of liquid and sludge, will be encapsulated by and dispersed in a borosilicate glass before permanent storage in a HLW repository. This glass containing the DHLW serves to dilute the radionuclides and to retard their dispersion into the environment. 318 refs

Natural glass analogues to alteration of nuclear waste glass: A review and recommendations for further study

Energy Technology Data Exchange (ETDEWEB)

McKenzie, W.F.

1990-01-01

The purpose of this report is to review previous work on the weathering of natural glasses; and to make recommendations for further work with respect to studying the alteration of natural glasses as it relates quantifying rates of dissolution. the first task was greatly simplified by the published papers of Jercinovic and Ewing (1987) and Byers, Jercinovic, and Ewing (1987). The second task is obviously the more difficult of the two and the author makes no claim of completeness in this regard. Glasses weather in the natural environment by reacting with aqueous solutions producing a rind of secondary solid phases. It had been proposed by some workers that the thickness of this rind is a function of the age of the glass and thus could be used to estimate glass dissolution rates. However, Jercinovic and Ewing (1987) point out that in general the rind thickness does not correlate with the age of the glass owing to the differences in time of contact with the solution compared to the actual age of the sample. It should be noted that the rate of glass dissolution is also a function of the composition of both the glass and the solution, and the temperature. Quantification of the effects of these parameters (as well as time of contact with the aqueous phase and flow rates) would thus permit a prediction of the consequences of glass-fluid interactions under varying environmental conditions. Defense high- level nuclear waste (DHLW), consisting primarily of liquid and sludge, will be encapsulated by and dispersed in a borosilicate glass before permanent storage in a HLW repository. This glass containing the DHLW serves to dilute the radionuclides and to retard their dispersion into the environment. 318 refs.

Physicochemical properties and long-term behavior of french R7T7 nuclear waste glass

International Nuclear Information System (INIS)

Vernaz, E.

1990-01-01

The French R7T7 nuclear glass composition was carefully selected to allow incorporation of some thirty different oxides found in fission product solutions. The resulting glass exhibits very low crystallization, and its physical and chemical properties are very similar to those of standard industrial glasses. Nuclear glasses have been shown to withstand α doses corresponding to several hundred thousand years under repository conditions. Predicting the long-term behavior of fission product glasses subjected to aqueous corrosion is no doubt the most difficult aspect of the problem. Predictions are necessarily based on mathematical models. A substantial research effort has been undertaken to identify all the basic corrosion mechanisms liable to control long-term alteration. These mechanisms are now relatively well understood, and provide the basis for developing the indispensable models. Realistic storage conditions exist under which glass alteration occurs at a very slow rate, and can fulfill its role as the first containment barrier for several tens of thousands of years

Study of rhyolitic glasses alteration in contact with natural brines (Bolivia). Application to the study of the long-term behaviour of the R7T7 nuclear glass

International Nuclear Information System (INIS)

Abdelouas, A.

1996-01-01

The purpose of this work is to complement an experimental program on the R7T7 nuclear waste glass alteration in brines at 190 deg C in Germany by the analysis of the structure and the chemical composition of the alteration layers, and to study the alteration of rhyolitic glasses in natural brines from Bolivia as analogue for nuclear waste glasses disposed in salt formations. Alteration experiments with the R7T7 and basaltic glasses and obsidian in MgCl 2 -CaCl 2 -saturated brine at 190 deg. C were also conducted in order to study the influence of the glass composition on the nature of the secondary phases. The experiments with the R7T7 glass in three salt brines, saturated respectively in MgCl 2 , MgCl 2 -CaCl 2 and NaCl, showed that the solubilities of most radionuclides are controlled by the secondary phases. Nd, La, and Pr are trapped in powellite, Ce in cerianite, U in coffinite, and Sr is partially immobilized in barite. These phases are stable for more than one year. There is a good similarity between the secondary phases formed experimentally on volcanic glasses and the R7T7 glass altered in MgCl 2 -CaCl 2 -saturated brine. The abundance of Mg in solution permits the formation of similar magnesian clays on the glass samples independently of the nature of the initial glasses. These results support the use of volcanic glasses alteration patterns in Mg-rich solutions to understand the long-term behavior of nuclear waste glasses and to evaluate the stability of the secondary phases. The study of the sediments of Uyuni (Bolivia) showed that the corrosion rate of the rhyolitic glass in brines at 10 deg. C is 12 to 30 time lower than those of rhyolitic glasses altered in high dilute conditions. The low alteration rate of rhyolitic glasses in brines and the formation of secondary phases such as smectite, barite and cerianite (also formed during the experimental alteration of the R7T7 glass), permit us to expect the low alteration of nuclear waste glasses at long

A comparison of the performance of nuclear waste glasses by modeling

International Nuclear Information System (INIS)

Grambow, B.; Strachan, D.M.

1988-01-01

A model selected for the licensing process must be based on a physical and chemical understanding of the glass corrosion mechanism. The purpose of this paper is to show that a dissolution/precipitation model can be used to better understand the effects of various system variables on glass dissolution. The application and validation of this model are also discussed. A dissolution/precipitation model developed appears applicable to experiments with a wide range of solution compositions as well as to more complex systems, such as the bentonite/glass/water system the steel corrosion product/glass/water system, or the dissolution of natural basalt glass in a geologic environment. This model is based on solution chemistry and transition state theory. The theoretical background of this model is discussed elsewhere and is used to describe the dissolution behavior of three nuclear waste glasses. These glasses were selected because they represent a wide range of behavior and, therefore, could be used to illustrate the capabilities of the dissolution/precipitation model. The effects of parameters, such as temperature and starting solution composition, on the dissolution behavior of glass are also discussed. 27 refs., 10 figs., 1 tab

XAFS study on silica glasses irradiated in a nuclear reactor

International Nuclear Information System (INIS)

Yoshida, Tomoko; Yoshida, Hisao; Hara, Takanobu; Ii, Tatsuya; Okada, Tomohisa; Tanabe, Tetsuo

2000-01-01

X-ray absorption technique (XANES and EXAFS) was applied to study the local structures of silica glasses before and after the irradiation in a nuclear reactor. Although our separate photoluminescence (PL) measurements clearly showed the different aspects about oxygen vacancies in these samples, i.e., at least the B 2β type oxygen-deficient center exists as an intrinsic defect in the fused silica glass while another type B 2α center is formed in the synthesized silica glass, such differences did not directly reflect on the X-ray absorption spectra (XANES and EXAFS). However, the curve-fitting analysis of EXAFS showed that the number of oxygen atoms coordinated to Si relatively increased after the irradiation. This result may indicate the occurrence of the structural relaxation in the irradiated samples, that is, a slightly distorted SiO 4 tetrahedra in silica glasses relaxed to the regular SiO 4 tetrahedra due to the break of some connections between SiO 4 units in the silica glasses. Thus, the X-ray absorption technique gave the important information of the in-reactor irradiated silica glasses which complements the results obtained from PL measurements

Progress toward bridging from atomistic to continuum modeling to predict nuclear waste glass dissolution.

Energy Technology Data Exchange (ETDEWEB)

Zapol, Peter (Argonne National Laboratory, Argonne, IL); Bourg, Ian (Lawrence Berkeley National Laboratories, Berkeley, CA); Criscenti, Louise Jacqueline; Steefel, Carl I. (Lawrence Berkeley National Laboratories, Berkeley, CA); Schultz, Peter Andrew

2011-10-01

This report summarizes research performed for the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Subcontinuum and Upscaling Task. The work conducted focused on developing a roadmap to include molecular scale, mechanistic information in continuum-scale models of nuclear waste glass dissolution. This information is derived from molecular-scale modeling efforts that are validated through comparison with experimental data. In addition to developing a master plan to incorporate a subcontinuum mechanistic understanding of glass dissolution into continuum models, methods were developed to generate constitutive dissolution rate expressions from quantum calculations, force field models were selected to generate multicomponent glass structures and gel layers, classical molecular modeling was used to study diffusion through nanopores analogous to those in the interfacial gel layer, and a micro-continuum model (K{mu}C) was developed to study coupled diffusion and reaction at the glass-gel-solution interface.

Study of powellite-rich glass-ceramics for nuclear waste immobilization

International Nuclear Information System (INIS)

Taurines, T.

2012-01-01

MoO 3 is poorly soluble in borosilicate glasses which can lead to the crystallization of undesired phases when its concentration or the charge load (minor actinides and fission products concentration) is too high. Crystallization control is needed to guarantee good immobilization properties. We studied powellite-rich glass-ceramics obtained from a simplified nuclear glass in the system SiO 2 - B 2 O 3 - Na 2 O - CaO - Al 2 O 3 - MoO 3 - RE 2 O 3 (RE = Gd, Eu, Nd) by various heat treatments. Rare earth elements (REE) were added as minor actinides surrogates and as spectroscopic probes. The influence of MoO 3 and RE 2 O 3 content on powellite (CaMoO 4 ) crystallization was investigated. Various glass-ceramics (similar residual glass + powellite) were obtained with large crystal size distributions. Phase separation due to molybdenum occurs during quenching when [MoO 3 ] ≥ 2.5 mol%. We showed that increasing the rare earth content can suppress the phase separation due to molybdenum but it leads to spinodal decomposition of the residual glass. Furthermore, we studied the effects of parent glass complexifying and the insertion of Gd 3+ ions into the powellite structure. In order to understand the influence of microstructure on evolutions under β-irradiation, we studied point defects creation and structural changes. We showed that the damage induced by electronic excitations in the glass-ceramics is driven by the damage in the residual glass. (author) [fr

A review of phase separation in borosilicate glasses, with reference to nuclear fuel waste immobilization

International Nuclear Information System (INIS)

Taylor, P.

1990-08-01

This report reviews information on miscibility limits in borosilicate glass-forming systems. It includes both a literature survey and an account of experimental work performed within the Canadian Nuclear Fuel Waste Management Program. Emphasis is placed on the measurement and depiction of miscibility limits in multicomponent (mainly quaternary) systems, and the effects of individual components on the occurrence of phase separation. The behaviour of the multicomponent system is related to that of simpler (binary and ternary) glass systems. The possible occurrence of phase separation, as well as its avoidance, during processing of nuclear waste glasses is discussed

An introduction to the impact of self-irradiation on nuclear glasses

International Nuclear Information System (INIS)

Ghaleb, D.

1997-01-01

Irradiation is an important factor in the aging of glasses used to confine nuclear wastes. Beta decays are prevailing during the first 500 years then alpha decays take over. Alpha decay is the main cause of atomic displacements while beta decay, through the emission of energetic electrons, is responsible for most electronic excitations and ionization. Each alpha decay is followed by the creation of an helium atom trapped inside the bulk of glass. (A.C.)

Thermodynamical modeling of nuclear glasses: coexistence of amorphous phases

International Nuclear Information System (INIS)

Adjanor, G.

2007-11-01

Investigating the stability of borosilicate glasses used in the nuclear industry with respect to phase separation requires to estimate the Gibbs free energies of the various phases appearing in the material. In simulation, using current computational resources, a direct state-sampling of a glassy system with respect to its ensemble statistics is not ergodic and the estimated ensemble averages are not reliable. Our approach consists in generating, at a given cooling rate, a series of quenches, or paths connecting states of the liquid to states of the glass, and then in taking into account the probability to generate the paths leading to the different glassy states in ensembles averages. In this way, we introduce a path ensemble formalism and calculate a Landau free energy associated to a glassy meta-basin. This method was validated by accurately mapping the free energy landscape of a 38-atom glassy cluster. We then applied this approach to the calculation of the Gibbs free energies of binary amorphous Lennard-Jones alloys, and checked the correlation between the observed tendencies to order or to phase separate and the computed Gibbs free energies. We finally computed the driving force to phase separation in a simplified three-oxide nuclear glass modeled by a Born-Mayer-Huggins potential that includes a three-body term, and we compared the estimated quantities to the available experimental data. (author)

The remote handling of canisters containing nuclear waste in glass at the Savannah River Plant

International Nuclear Information System (INIS)

Callan, J.E.

1986-01-01

The Defense Waste Processing Facility (DWPF) is a complete production area being constructed at the Savannah River Plant for the immobilization of nuclear waste in glass. The remote handling of canisters filled with nuclear waste in glass is an essential part of the process of the DWPF at the Savannah River Plant. The canisters are filled with nuclear waste containing up to 235,000 curies of radioactivity. Handling and movement of these canisters must be accomplished remotely since they radiate up to 5000 R/h. Within the Vitrification Building during filling, cleaning, and sealing, canisters are moved using standard cranes and trolleys and a specially designed grapple. During transportation to the Glass Waste Storage Building, a one-of-a-kind, specially designed Shielded Canister Transporter (SCT) is used. 8 figs

Effect of irradiation on the evolution of alteration layer formed during nuclear glass leaching

International Nuclear Information System (INIS)

Mougnaud, Sarah

2016-01-01

High-level radioactive waste (HLW) remaining after spent nuclear fuel reprocessing is immobilized within a glass matrix, eventually destined for geological disposal. Water intrusion into the repository is expected after several thousand years. The alteration of a non-radioactive surrogate for nuclear glass has been extensively studied and it has been determined that successive leaching mechanisms lead to the formation of a 'passivating' alteration layer and to the establishment of a residual rate regime in the long term. However, glass packages are submitted to the radioactivity of confined radioelements. This work focuses on the influence of irradiation on the alteration layer formed during the residual rate regime, in a structural and mechanistic point of view. Three focal areas have been selected. Non-radioactive simple glasses have been leached and externally irradiated in order to determine modifications induced by electronic effects (irradiations with electrons and alpha particles). The same type of glass samples have been previously irradiated with heavy ions and their leaching behavior have been studied in order to assess the impact of ballistic dose cumulated by the glass before water intrusion. Leaching behavior of a complex radioactive glass, doped with an alpha-emitter, has been studied to consider a more realistic situation. (author) [fr

A comparison of the performance of nuclear waste glasses by modeling

International Nuclear Information System (INIS)

Grambow, B.; Strachan, D.M.

1988-12-01

Through a combination of data collection and computer modeling, the dissolution mechanism of nuclear waste glasses has been investigated and more clearly defined. Glass dissolution can be described as a dissolution/precipitation process in which glass dissolves in aqueous solution and solids precipitate as the solubility products are exceeded. The dissolution process is controlled by activity of the rate-limiting specie H 4 SiO 4 . As a concentration of H 4 SiO 4 increases, the rate of dissolution decreases until a final reaction rate is reached. Between the forward reaction rate (early time) and final reaction rate (very long time), glasses may exhibit an intermediate root time dependence caused by a transport resistance for the diffusion of H 4 SiO 4 within the gel layer on the glass surface. In this report, three glasses are studied: JSS-A, PNL 76-68, and SRL-131. Data from static and dynamic leach tests are assembled, plotted, and successfully modeled. The kinetic parameters for these glasses are reported. With four parameters derived from experiments for each glass, the model can be used to calculate the effects of changes in the initial composition of the water contacting the glass. The effects of convective flow can also be modeled. Furthermore, glasses of different compositions can be readily compared. 49 refs., 27 figs., 5 tabs

The effect of chromium oxide on the properties of simulated nuclear waste glasses

International Nuclear Information System (INIS)

Vojtech, O.; Sussmilch, J.; Urbanec, Z.

1996-02-01

A study of the effect of chromium on the properties of selected glasses was performed in the frame of a Contract between Battelle, Pacific Northwest Laboratories and Nuclear Research Institute, ReZ. In the period from July 1994 to June 1995 two borosilicate glasses of special composition were prepared according to the PNL procedure and their physical and structural characteristics of glasses were studied. This Final Report contains a vast documentation on the properties of all glasses studied. For the preparation of the respective technology more detailed study of physico-chemical properties and crystallinity of investigated systems would be desirable

Vanadium and Chromium Redox Behavior in borosilicate Nuclear Waste Glasses

International Nuclear Information System (INIS)

McKeown, D.; Muller, I.; Gan, H.; Feng, Z.; Viragh, C.; Pegg, I.

2011-01-01

X-ray absorption spectroscopy (XAS) was used to characterize vanadium (V) and chromium (Cr) environments in low activity nuclear waste (LAW) glasses synthesized under a variety of redox conditions. V 2 O 5 was added to the melt to improve sulfur incorporation from the waste; however, at sufficiently high concentrations, V increased melt foaming, which lowered melt processing rates. Foaming may be reduced by varying the redox conditions of the melt, while small amounts of Cr are added to reduce melter refractory corrosion. Three parent glasses were studied, where CO-CO 2 mixtures were bubbled through the corresponding melt for increasing time intervals so that a series of redox-adjusted-glasses was synthesized from each parent glass. XAS data indicated that V and Cr behaviors are significantly different in these glasses with respect to the cumulative gas bubbling times: V 4+ /V total ranges from 8 to 35%, while Cr 3+ /Cr total can range from 15 to 100% and even to population distributions including Cr 2+ . As Na-content decreased, V, and especially, Cr became more reduced, when comparing equivalent glasses within a series. The Na-poor glass series show possible redox coupling between V and Cr, where V 4+ populations increase after initial bubbling, but as bubbling time increases, V 4+ populations drop to near the level of the parent glass, while Cr becomes more reduced to the point of having increasing Cr 2+ populations.

Molecular dynamics study of a nuclear waste glass matrix with plutonium

International Nuclear Information System (INIS)

Meis, C.; Delaye, J.M.; Ghaleb, D.

1999-01-01

Molecular dynamics simulation techniques were applied to model the incorporation of plutonium in the French nuclear waste glass matrix. Born-Mayer-Huggins analytical potentials were established to characterize short-range interactions between Pu-O and Pu-Pu pairs; the potentials were fitted to the structural properties of plutonium dioxide in the light of a recent experimental study showing that plutonium is found as Pu(IV) in the glass. The transferability of the established potentials to the glass structure is discussed, and the potential parameters are further refined by molecular dynamics simulations in an aluminoborosilicate glass to obtain mean Pu-O interatomic distances and first-neighbor coordination numbers matching the experimental values as closely as possible. Previously published Born-Mayer-Huggins potentials supplemented by Stillinger-Weber three-body terms were used for oxygen-cation and cation-cation interactions. The difficulties encountered in establishing a Pu-O potential that provides satisfactory results in both oxides and glasses are also discussed

Deformation mechanisms during nanoindentation of sodium borosilicate glasses of nuclear interest

Energy Technology Data Exchange (ETDEWEB)

Kilymis, D. A.; Delaye, J.-M., E-mail: jean-marc.delaye@cea.fr [CEA Marcoule, DEN/DTCD, Service d’Etude et Comportement des Matériaux de Conditionnement, BP17171 30207 Bagnols-sur-Cèze Cedex (France)

2014-07-07

In this paper we analyze results of Molecular Dynamics simulations of Vickers nanoindentation, performed for sodium borosilicate glasses of interest in the nuclear industry. Three glasses have been studied in their pristine form, as well as a disordered one that is analogous to the real irradiated glass. We focused in the behavior of the glass during the nanoindentation in order to reveal the mechanisms of deformation and how they are affected by microstructural characteristics. Results have shown a strong dependence on the SiO{sub 2} content of the glass, which promotes densification due to the open structure of SiO{sub 4} tetrahedra and also due to the strength of Si-O bonds. Densification for the glasses is primarily expressed by the relative decrease of the Si-O-Si and Si-O-B angles, indicating rotation of the structural units and decrease of free volume. The increase of alkali content on the other hand results to higher plasticity of the matrix and increased shear flow. The most important effect on the deformation mechanism of the disordered glasses is that of the highly depolymerized network that will also induce shear flow and, in combination with the increased free volume, will result in the decreased hardness of these glasses, as has been previously observed.

Deformation mechanisms during nanoindentation of sodium borosilicate glasses of nuclear interest.

Science.gov (United States)

Kilymis, D A; Delaye, J-M

2014-07-07

In this paper we analyze results of Molecular Dynamics simulations of Vickers nanoindentation, performed for sodium borosilicate glasses of interest in the nuclear industry. Three glasses have been studied in their pristine form, as well as a disordered one that is analogous to the real irradiated glass. We focused in the behavior of the glass during the nanoindentation in order to reveal the mechanisms of deformation and how they are affected by microstructural characteristics. Results have shown a strong dependence on the SiO2 content of the glass, which promotes densification due to the open structure of SiO4 tetrahedra and also due to the strength of Si-O bonds. Densification for the glasses is primarily expressed by the relative decrease of the Si-O-Si and Si-O-B angles, indicating rotation of the structural units and decrease of free volume. The increase of alkali content on the other hand results to higher plasticity of the matrix and increased shear flow. The most important effect on the deformation mechanism of the disordered glasses is that of the highly depolymerized network that will also induce shear flow and, in combination with the increased free volume, will result in the decreased hardness of these glasses, as has been previously observed.

Natural analogue of nuclear waste glass in a geologic formation. Study on long-term behavior of volcanic glass shards collected from drill cores

International Nuclear Information System (INIS)

Yoshikawa, Hideki; Yui, Mikazu; Futakuchi, Katsuhito; Hiroki, Minenari

2005-01-01

Alteration of the volcanic glass in geologic formation was investigated as one of the natural analog for a glass of high-level nuclear waste in geological disposal. We analyzed some volcanic glasses included in the core sample of the bore hole and estimated the history of its burying and observed its alteration using the polarizing microscope. Some information at the piling up temperature and the piling up time was collected. (author)

Effect of Na{sub 2}O on aqueous dissolution of nuclear waste glasses

Energy Technology Data Exchange (ETDEWEB)

Farooqi, Rahmat Ullah, E-mail: rufarooqi@live.com [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology, 77 Cheongam-Ro, Nam-Gu, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Hrma, Pavel [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology, 77 Cheongam-Ro, Nam-Gu, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Pacific Northwest National Laboratory, Richland, WA (United States)

2017-04-15

Sodium oxide is present in the majority of commercial and waste glasses as a viscosity-reducing component. In some nuclear waste glasses, its source is the waste itself. As such, it can limit the waste loading because of its deleterious effect on the resistance of the glass to attack by aqueous media. The maximum tolerable content of Na{sub 2}O in glass depends on the presence and concentration of components that interact with it. To assess the acceptability limits of Na{sub 2}O in the composition region of nuclear waste glasses, we formulated 11 baseline compositions by varying the content of oxides of Si, B, Al, Ca, Zr, and Li. In each of these compositions, we varied the Na{sub 2}O fraction from 8–16 mass% to 23–30 mass%. To each of 146 glasses thus formulated, we applied the seven-day Product Consistency Test (PCT) to determine normalized B and Na releases (r{sub i}, where i ≡ B or Na). Fitting approximation functions ln(r{sub i}/gm{sup −2}) = Σb{sub ij}g{sub j} to r{sub i} data (g{sub j} is the j-th component mass fraction and b{sub ij} the corresponding component coefficient), we showed that the r{sub B} (and, consequently, the initial glass alteration rate) was proportional to the glass component mass fractions in the order Al{sub 2}O{sub 3}glass structure would fall apart or beyond which a continuous nondurable phase would be separated. Specific examples are given to demonstrate restrictions imposed on the boundary of the composition region of acceptable glasses by the maximum allowable r{sub B} and by the melt viscosity required for glass melter operation. Finally, the role that PCT data may play in understanding the evolution of the glass alteration process is discussed.

High-level nuclear waste borosilicate glass: A compendium of characteristics

International Nuclear Information System (INIS)

Cunnane, J.C.; Bates, J.K.; Ebert, W.L.; Feng, X.; Mazer, J.J.; Wronkiewicz, D.J.; Sproull, J.; Bourcier, W.L.; McGrail, B.P.

1992-01-01

With the imminent startup, in the United States, of facilities for vitrification of high-level nuclear waste, a document has been prepared that compiles the scientific basis for understanding the alteration of the waste glass products under the range of service conditions to which they may be exposed during storage, transportation, and eventual geologic disposal. A summary of selected parts of the content of this document is provided. Waste glass alterations in a geologic repository may include corrosion of the glass network due to groundwater and/or water vapor contact. Experimental testing results are described and interpreted in terms of the underlying chemical reactions and physical processes involved. The status of mechanistic modeling, which can be used for long-term predictions, is described and the remaining uncertainties associated with long-term simulations are summarized

Influence of Some Nuclear Waste on The Durability and Mechanical Properties of Borosilicate glass

International Nuclear Information System (INIS)

El-Alaily, N.A.

2003-01-01

Various glass systems have been shown to be suitable for producing waste glass forms that are thermally and mechanically stable and exhibit good chemical durability. In this study borosilicate glass containing sodium oxide and aluminum oxide was prepared as a host for high level nuclear waste. The glass durability when the samples were immersed either in distilled water or ground water at 70 degree was studied. The density, porosity and mechanical properties were also investigated. The effects of exposing the samples immersed in groundwater to gamma rays in the glass durability and all other mentioned properties were also studied. The results showed that immersing the glass in ground water causing a decrease in the glass durability. The exposure of the glass immersed in ground water to the gamma rays increases the durability of the glass. The mechanical properties of the prepared glass were good. Although these properties decrease for the corroded glass but they were still good

Relationship between reaction layer thickness and leach rate for nuclear waste glasses

International Nuclear Information System (INIS)

Chick, L.A.; Pederson, L.R.

1984-02-01

Three leaching tests, devised to distinguish among several proposed nuclear waste glass leaching mechanisms, were carried out for four different waste glasses. In the first test, the influence of a pre-formed reaction layer on elemental release was evaluated. In the second test, glass specimens were replaced with fresh samples halfway through the leaching experiment, to evaluate the influence of the concentration of glass components in leaching. Finally, regular replacement of the leachant at fixed time intervals essentially removed the variable changing solution concentration, and allowed an assessment of the influence of reaction layer thickness on the leaching rate. Results for all glasses tested indicated that the reaction layer presented little or no barrier to leaching, and that most of the retardation on leaching rates generally observed are attributable to saturation effects. 20 references, 6 figures, 1 table

Simulation of Self-Irradiation of High-Sodium Content Nuclear Waste Glasses

International Nuclear Information System (INIS)

Pankov, Alexey S.; Ojovan, Michael I.; Batyukhnova, Olga G.; Lee, William E.

2007-01-01

Alkali-borosilicate glasses are widely used in nuclear industry as a matrix for immobilisation of hazardous radioactive wastes. Durability or corrosion resistance of these glasses is one of key parameters in waste storage and disposal safety. It is influenced by many factors such as composition of glass and surrounding media, temperature, time and so on. As these glasses contain radioactive elements most of their properties including corrosion resistance are also impacted by self-irradiation. The effect of external gamma-irradiation on the short-term (up to 27 days) dissolution of waste borosilicate glasses at moderate temperatures (30 deg. to 60 deg. C) was studied. The glasses studied were Magnox Waste glass used for immobilisation of HLW in UK, and K-26 glass used in Russia for ILW immobilisation. Glass samples were irradiated under γ-source (Co-60) up to doses 1 and 11 MGy. Normalised rates of elemental release and activation energy of release were measured for Na, Li, Ca, Mg, B, Si and Mo before and after irradiation. Irradiation up to 1 MGy results in increase of leaching rate of almost all elements from both MW and K-26 with the exception of Na release from MW glass. Further irradiation up to a dose of 11 MGy leads to the decrease of elemental release rates to nearly initial value. Another effect of irradiation is increase of activation energies of elemental release. (authors)

Mechanisms that control aqueous leaching of nuclear waste glass

International Nuclear Information System (INIS)

Simmons, J.H.; Barkatt, A.; Macedo, P.B.

1982-01-01

The development of predictive models and risk calculations for the time evolution of radioactive isotope leaching from fixation solids depends on many factors, including measurement accuracy, measurement relevance, a complete understanding of possible dissolution mechanisms, and the ability to project worst case conditions for all appropriate mechanisms. Some of the mechanisms observed and understood at present are the mechanisms of dissolution of glasses in neutral unbuffered water, the effects of structural disintegration of the glass protective layer, the effects of slowly flowing bath waters and possibly, the mechanisms of leach-rate reduction by solution saturation. The mechanisms that control radiation and temperature effects including alpha particle emission and nuclear transmutations are as yet little understood or investigated

A review of glass-ceramics for the immobilization of nuclear fuel recycle wastes

International Nuclear Information System (INIS)

Hayward, P.J.

1987-01-01

This report reviews the status of the Canadian, German, U.S., Japanese, U.S.S.R. and Swedish programs for the development of glass-ceramic materials for immobilizing the high-level radioactive wastes arising from the recycling of used nuclear fuel. The progress made in these programs is described, with emphasis on the Canadian program for the development of sphene-based glass-ceramics. The general considerations of product performance and process feasibility for glass-ceramics as a category of waste form material are discussed. 137 refs

Mineralogy and thermodynamic properties of magnesium phyllosilicates formed during the alteration of a simplified nuclear glass

Energy Technology Data Exchange (ETDEWEB)

Debure, Mathieu, E-mail: m.debure@brgm.fr [CEA, DEN, DTCD, SECM, F-30207 Bagnols-sur-Ceze (France); MINES-ParisTech, PSL Research University, Centre de Géosciences, 77305 Fontainebleau (France); De Windt, Laurent [MINES-ParisTech, PSL Research University, Centre de Géosciences, 77305 Fontainebleau (France); Frugier, Pierre; Gin, Stéphane [CEA, DEN, DTCD, SECM, F-30207 Bagnols-sur-Ceze (France); Vieillard, Philippe [IC2MP-CNRS-UMR 7285, 5 Ave. Albert Turpain TSA 51106, 86073 Poitiers Cedex 09 (France)

2016-07-15

The precipitation of crystallized magnesium phyllosilicates generally sustains the alteration rate of nuclear waste containment glass. However, glass alteration slows down to a residual rate as soon as Mg disappears from the solution. The identification of the phyllosilicates formed is therefore crucial for modeling the long-term behavior of nuclear glass. This study deals with batch alteration of the simplified nuclear glass ISG in presence of magnesium, and the characterization of the secondary phases. Morphological, chemical and structural analyses (MET, EDX, XRD) were performed to determine the nature and structure of the precipitated phases identified as trioctahedral smectites. Analyses conducted on the secondary phases proved the presence of Al, Na and Ca in the Mg-phyllosilicate phases. Such elements had been suspected but never quantitatively measured. The experimental results were then used to determine the thermodynamic solubility constants for each precipitated secondary phase at various temperatures. The calculated values were consistent with those available for sodium and magnesium saponites in the existing thermodynamic databases. - Highlights: • The international simple glass dissolution rate increases in presence of magnesium. • Mg added in solution combines with Si from glass to yield trioctahedral smectites. • Their calculated logK are close to smectite thermodynamic constants reported in databases. • It confirms assumptions on Mg-silicates phases made in previous geochemical modeling.

The encapsulation of nuclear waste in a magnesium aluminosilicate glass-ceramic

International Nuclear Information System (INIS)

Luk, K.M.

1999-07-01

The use of Magnesium aluminosilicate (MAS) glass-ceramics for the immobilisation of nuclear waste has been investigated. Nuclear waste is currently immobilised in a borosilicate glass. It is possible that immobilisation in an MAS glass-ceramic will reduce processing temperature of the waste, offer greater thermal and chemical stabilities and chemical durabilities. The primary reason for investigating sintered glass-ceramics is the possible advent of wastes containing high levels of refractory elements such as zirconia from the future reprocessing techniques such as electrochemical dissolution. In the first instance zirconia was used as a simulated waste with the principal of encapsulating zirconia with the minimum of porosity. Attempts were made to encapsulate 0, 20 and 40 volume % of zirconia in MAS sintering at temperatures of around 950 deg. C. It was found that the main cause of porosity was the agglomeration of fine zirconia powder. Three Taguchi experiments to optimise conditions for encapsulation of zirconia in MAS were carried out. In each case 10 volume % of zirconia was encapsulated. A Taguchi L 8 was carried out to optimise thermal conditions and powder characteristics. A Taguchi L 9 was carried out to improve knowledge of the thermal characteristics and an L 16 was carried out to provide information on curvature of thermal parameters and powder particle sizes. The conditions predicted to be optimum from these Taguchi experiments were a temperature of 940 - 960 deg. C, a heating rate of 30 deg. C/min, a hold time of 30 - 50 minutes and particle sizes of 2-4 and ∼ 15μm respectively. Densifications of up to 99% have been observed. Tapping experiments were carried out in an attempt to remove the pressing stage from processing. MAS was tapped into an alumina crucible with and without the addition of a dead weight. Almost fully dense MAS pellets were produced. This is an indication that it may be possible to process glass-ceramic waste forms in their final

Effects of deposited nuclear and electronic energy on the hardness of R7T7-type containment glass

Energy Technology Data Exchange (ETDEWEB)

Peuget, S. [Commissariat a l' Energie Atomique, CEA Marcoule, DEN/DTCD/SECM/LMPA, Batiment 166, BP 17171, F-30207 Bagnols-sur-Ceze Cedex (France)]. E-mail: sylvain.peuget@cea.fr; Noel, P.-Y. [Commissariat a l' Energie Atomique, CEA Marcoule, DEN/DTCD/SECM/LMPA, Batiment 166, BP 17171, F-30207 Bagnols-sur-Ceze Cedex (France); Loubet, J.-L. [Laboratoire de Tribologie et Dynamique des Systemes, UMR CNRS 5513, Ecole Centrale de Lyon 36, avenue Guy de Collongue, 69134 Ecully Cedex (France); Pavan, S. [Laboratoire de Tribologie et Dynamique des Systemes, UMR CNRS 5513, Ecole Centrale de Lyon 36, avenue Guy de Collongue, 69134 Ecully Cedex (France); Nivet, P. [Commissariat a l' Energie Atomique, CEA Marcoule, DEN/DTCD/SECM/LMPA, Batiment 166, BP 17171, F-30207 Bagnols-sur-Ceze Cedex (France); Chenet, A. [Commissariat a l' Energie Atomique, CEA Marcoule, DEN/DTCD/SECM/LMPA, Batiment 166, BP 17171, F-30207 Bagnols-sur-Ceze Cedex (France)

2006-05-15

The effects of elastic and inelastic interactions induced by cumulative alpha decay on the hardness of R7T7-type nuclear containment glass were investigated on actinide-doped glass specimens and by external irradiation of inactive glass by light and heavy ions. Vickers microindentation and nanoindentation hardness measurements showed that in the deposited energy range investigated (below 3 x 10{sup 22} keV/cm{sup 3}) inelastic effects have no influence on the plastic response of the glass. Conversely, identical hardness variations versus the nuclear energy deposited in the material were observed on curium-doped glass and on glass irradiated by ion bombardment. The observed hardness variation stabilized after the deposited energy reached about 3 x 10{sup 2} keV{sub nucl}/cm{sup 3}. These findings indicate that the change in the plastic response of the glass is a consequence of ballistic effects.

Materials interactions relating to long-term geologic disposal of nuclear waste glass

International Nuclear Information System (INIS)

Bibler, N.E.; Jantzen, C.M.

1987-01-01

In the geologic disposal of nuclear waste glass, the glass will eventually interact with groundwater in the repository system. Interactions can also occur between the glass and other waste package materials that are present. These include the steel canister that holds the glass, the metal overpack over the canister, backfill materials that may be used, and the repository host rock. This review paper systematizes the additional interactions that materials in the waste package will impose on the borosilicate glass waste form-groundwater interactions. The repository geologies reviewed are tuff, salt, basalt, and granite. The interactions emphasized are those appropriate to conditions expected after repository closure, e.g. oxic vs anoxic conditions. Whenever possible, the effect of radiation from the waste form on the interactions is examined. The interactions are evaluated based on their effect on the release and speciation of various elements including radionuclides from the glass. It is noted when further tests of repository interactions are needed before long-term predictions can be made. 63 references, 1 table

Standard test methods for determining chemical durability of nuclear, hazardous, and mixed waste glasses and multiphase glass ceramics: The product consistency test (PCT)

CERN Document Server

American Society for Testing and Materials. Philadelphia

2002-01-01

1.1 These product consistency test methods A and B evaluate the chemical durability of homogeneous glasses, phase separated glasses, devitrified glasses, glass ceramics, and/or multiphase glass ceramic waste forms hereafter collectively referred to as “glass waste forms” by measuring the concentrations of the chemical species released to a test solution. 1.1.1 Test Method A is a seven-day chemical durability test performed at 90 ± 2°C in a leachant of ASTM-Type I water. The test method is static and conducted in stainless steel vessels. Test Method A can specifically be used to evaluate whether the chemical durability and elemental release characteristics of nuclear, hazardous, and mixed glass waste forms have been consistently controlled during production. This test method is applicable to radioactive and simulated glass waste forms as defined above. 1.1.2 Test Method B is a durability test that allows testing at various test durations, test temperatures, mesh size, mass of sample, leachant volume, a...

Structure and properties of rare earth-rich glassed for nuclear waste immobilisation

International Nuclear Information System (INIS)

Bardez, I.

2004-11-01

A new nuclear glass composition, able to immobilize highly radioactive liquid wastes from high burn-up UO 2 fuel, was established and its structure studied. The composition of the selected rare earth-rich glass is (molar %): 61.79 SiO 2 - 8.94 B 2 O 3 - 3.05 Al 2 O 3 - 14.41 Na 2 O - 6.32 CaO - 1.89 ZrO 2 - 3.60 RE 2 O 3 (with RE = La, Ce, Pr and Nd). The aim of this study was to determine the local environment of the rare earth in this glass and also to glean information about the effect of glass composition on the rare earth neighbouring (influence of Si, B, Al, Na and Ca contents). To this end, several series of glasses, prepared from the baseline glass, were studied by different characterisation methods such as EXAFS spectroscopy at the neodymium LIII-edge, optical absorption spectroscopy, Raman spectroscopy and 29 Si, 27 Al and 11 B MAS-NMR. By coupling all the results obtained, several hypotheses about the nature of the rare earth neighbouring in the glass were proposed. (author)

PIXE and light element analysis (C,N) in glass inclusions trapped in meteorites with the nuclear microprobe

International Nuclear Information System (INIS)

Varela, M.E.; Mosbah, M.; Metrich, N.; Duraud, J.P.; Kurat, G.

1999-01-01

Proton-induced X-ray emission (PIXE) and light element analysis have been performed with the nuclear microprobe at the Laboratoire Pierre Suee (Saclay-France) in glass inclusions of the carbonaceous chondrites: Allende, Kaba and Renazzo, and in the achondrite meteorite: Chassigny. Carbon contents in olivine of chondrules are below the nuclear reactions analysis (NRA) detection limit, however, glasses from glass inclusions hosted by these grains, contain an appreciable and highly variable quantities of carbon (200-1600 ppm). This could indicate variable amounts of C trapped during glass inclusion formation. On the other hand, nitrogen is present in highly variable amounts in glasses of both, chondrites and achondrites minerals. Its abundance, correlated with depth from the section surface which suggests loss of N during analyses and therefore the possible existence of a very mobile (volatile?) species. A chondritic Rb/Sr and K/Rb ratio obtained by PIXE analyses in the glass-bearing inclusions of the Chassigny meteorite points towards a primitive source for the glass precursor of Chassigny inclusions

Chemical durability of borosilicate glasses containing simulated high-level nuclear wastes, 1

International Nuclear Information System (INIS)

Hara, Shigeo; Terai, Ryohei; Yamanaka, Hiroshi

1983-01-01

The Soxhlet-type leaching test apparatus has been developed to evaluate the chemical durability of some borosilicate glasses containing simulated High-Level nuclear Wastes, HLW. After the leaching over the temperature range of 50 0 -95 0 C, the weight loss of specimens with time was determined on both the samples of blocks and grains, and various components dissolved into water were analyzed by atomic absorption and colorimetry technique. It was found that Soxhlet-type test method was more useful than JIS test method, because the specimens in Soxhlet type apparatus were forced always to react with pure water and the mechanism of leaching could be evaluate accurately. The chemical durability of commercial glasses decreases generally with increasing of alkali contents in glasses. In the case of these borosilicate glasses containing HLW, however, the leachability was apparently independent on the alkali contents because of the complexity of these glass compositions. The variation of leaching rate with temperature suggests that dissolution mechanism changes with temperature. (author)

Water penetration mechanisms in nuclear glasses by X-ray and neutron reflectometry

International Nuclear Information System (INIS)

Rebiscoul, D.; Rieutord, F.; Ne, F.; Frugier, P.; Gin, S.; Cubitt, R.

2007-01-01

To determine the water diffusion at the early stage of the alteration, X-ray and neutron reflectometry have been performed on altered simplified glasses and the SON68 glass (an inactive R7T7-type French nuclear glass). For the first experiment, the simplified and SON68 glasses were altered at pH 3 and pH 6 and characterized by X-ray reflectometry as a function of the alteration duration. The evolutions of the electron density profile obtained from the reflectivity curves simulations have allowed the determination of the layers compositions. At the beginning of the alteration and for pH 3, the altered surface layer is constituted of a dealkalized zone. Upon alteration progress, the water diffuses inside the layer and hydrolyzes the Si-O-B bonds. For the second experiment, glasses were altered in D 2 O (pD 3) and analyzed in D 2 O saturated cell. After a D 2 O/H 2 O substitution, the samples were characterized one more time in H 2 O saturated cell. The evolution of the scattering length density shows that in the first stage of the alteration, the layer is constituted of two parts: a dealkalized glass and a dealkalized and boron depleted glass where water has diffused. According to the glass composition and after few hours of alteration, this dealkalized glass part can disappear. (authors)

Platinoids and molybdenum in nuclear waste containment glasses: a structural study

International Nuclear Information System (INIS)

Le Grand, M.

2000-01-01

This work deals with the structure of borosilicate nuclear glasses and with some relationships between structure and macroscopic properties. Two types of elements which may disturb the industrial process - platinoids (Ru and Pd) and molybdenum - are central to this work. Platinoids induce weak modifications on the structure of the glass, causing a depolymerization of the glassy network, an increase of the [3] B/ [4] B ratio and a modification of the medium range order around Si between 3.3 and 4.5 angstrom. The modifications of viscosity and density induced by platinoids in the glass are not due to the structural effect of the platinoids. The increase of viscosity is attributed to needle shaped RuO 2 . It can be moderated by imposing reducing conditions during the elaboration of the glass. The slight difference between experimental and calculated densities is due to the increase of the volume percentage of bubbles in the glass with increasing platinoid content. Mo is either present in the glass as molybdic groupings, or mobilized in chemically complex molybdic crystalline phases. The chemical composition and mineralogy of these phases has been obtained using electronic microprobe data and XRD with Rietveld analysis. The distribution of the different elements between the crystalline phases and the glass is strongly influenced by the structural role of the various cations in the glass. The Mo present in the glass appears as MoO 4 tetrahedra, independent of the borosilicate network. The formation of the crystalline phases can be explained by the existence of a precursor in which the MoO 4 tetrahedra are concentrated in rich alkali and earth-alkali bearing areas of the glass. (author)

Effects of β-irradiation in multicomponent glasses simulating the matrix of the French nuclear waste glass (R7T7)

International Nuclear Information System (INIS)

Boizot, B.; Ghaleb, D.; Petite, G.

2001-01-01

4-, 5- and 6-oxide components alumino-borosilicate glasses, with compositions closed to the matrix of the french nuclear glass 'R7T7' have been irradiated with electrons (β) at 2.5 MeV with a Van de Graff accelerator. These glasses have been studied after irradiation with different spectroscopic methods: Electron Paramagnetic Resonance for the study of defects, Raman Micro-spectroscopy for the study of amorphous network evolution under irradiation, and by 11 B MAS NMR. The results of these studies are presented here. It shows in particular a great sensibility to the irradiation conditions like dose rate and irradiation temperature, who are therefore important parameters for the representativeness of such experiments. (authors)

Thermal and physicochemical properties important for the long term behavior of nuclear waste glasses

International Nuclear Information System (INIS)

Vernaz, E.; Matzke, H.J.

1992-01-01

High level nuclear waste from reprocessing of spent nuclear fuel has to be solidified in a stable matrix for safe long-time storage. Vitrification in borosilicate glasses is the technique accepted worldwide. A number of different glasses was developed in different national programs. The criteria and the reasons for selecting the final compositions are briefly described. Emphasis is placed on the French product R7T7 and on thermal and physicochemical properties though glasses developed in other national projects (e.g. the German product GP 98/12 etc.) are also treated. The basic physical and mechanical properties and the chemical durability of the glass in contact with water or other aqueous solutions are described. The basic mechanisms of aqueous corrosion are discussed and the evolving modelling of the leaching process is dealt with, as well as effects of container material, backfill, etc. The thermal behavior has also been studied and extensive data exist on diffusion of glass constituents (Na) and of interesting elements of the waste such as the alkalis Rb and Cs or the actinides U and Pu, as well as on crystallization processes in the glass during storage at elevated temperatures. Emphasis is placed on the radiation stability of the glasses, based on extensive studies using short-lived actinides (e.g. Cm-244) or ion-implantation to produce the damage expected during long storage at an accelerated rate. The radiation stability is shown to be very good, if realistic damage conditions are used. The knowledge accumulated in the past years is used to evaluate and predict the long-term evolution of the glass under storage conditions

Effects of composition on properties in an 11-component nuclear waste glass system

International Nuclear Information System (INIS)

Chick, L.A.; Piepel, G.F.; Mellinger, G.B.; May, R.P.; Gray, W.J.; Buckwalter, C.Q.

1981-09-01

Ninety simplified nuclear waste glass compositions within an 11-component oxide composition matrix were tested for crystallinity, viscosity, volatility, and chemical durability. Empirical models of property response as a function of glass composition were developed using statistical experimental design and modeling techniques. A new statistical technique was developed to calculate the effects of oxide components on each property. Independent melts were used to check the prediction accuracy of the models

Compositional Dependence of Solubility/Retention of Molybdenum Oxides in Aluminoborosilicate-Based Model Nuclear Waste Glasses.

Science.gov (United States)

Brehault, Antoine; Patil, Deepak; Kamat, Hrishikesh; Youngman, Randall E; Thirion, Lynn M; Mauro, John C; Corkhill, Claire L; McCloy, John S; Goel, Ashutosh

2018-02-08

Molybdenum oxides are an integral component of the high-level waste streams being generated from the nuclear reactors in several countries. Although borosilicate glass has been chosen as the baseline waste form by most of the countries to immobilize these waste streams, molybdate oxyanions (MoO 4 2- ) exhibit very low solubility (∼1 mol %) in these glass matrices. In the past three to four decades, several studies describing the compositional and structural dependence of molybdate anions in borosilicate and aluminoborosilicate glasses have been reported in the literature, providing a basis for our understanding of fundamental science that governs the solubility and retention of these species in the nuclear waste glasses. However, there are still several open questions that need to be answered to gain an in-depth understanding of the mechanisms that control the solubility and retention of these oxyanions in glassy waste forms. This article is focused on finding answers to two such questions: (1) What are the solubility and retention limits of MoO 3 in aluminoborosilicate glasses as a function of chemical composition? (2) Why is there a considerable increase in the solubility of MoO 3 with incorporation of rare-earth oxides (for example, Nd 2 O 3 ) in aluminoborosilicate glasses? Accordingly, three different series of aluminoborosilicate glasses (compositional complexity being added in a tiered approach) with varying MoO 3 concentrations have been synthesized and characterized for their ability to accommodate molybdate ions in their structure (solubility) and as a glass-ceramic (retention). The contradictory viewpoints (between different research groups) pertaining to the impact of rare-earth cations on the structure of aluminoborosilicate glasses are discussed, and their implications on the solubility of MoO 3 in these glasses are evaluated. A novel hypothesis explaining the mechanism governing the solubility of MoO 3 in rare-earth containing aluminoborosilicate

Glass and nuclear wastes

International Nuclear Information System (INIS)

Sombret, C.

1982-10-01

Glass shows interesting technical and economical properties for long term storage of solidified radioactive wastes by vitrification or embedding. Glass composition, vitrification processes, stability under irradiation, thermal stability and aqueous corrosion are studied [fr

Effects of nuclear radiation on the optical properties of cerium-doped glass (accelerator applications)

CERN Document Server

McGrath, B; Van de Voorde, M H

1976-01-01

Some twenty types of glass containing 0.5-4% CeO/sub 2/ have been irradiated in a /sup 60/Co gamma cell and in the mixed neutron-gamma field of a nuclear reactor, at total integrated doses of up to 5*10 /sup 9/ rad (CH). The resulting colouration has been assessed quantitatively by measuring the light transmission with reference to air, in the range 360-510 nm. From the results, certain types of glass suitable for applications in nuclear engineering can be selected. Specifically, it was found that 1-2% CeO/sub 2/ content is usually sufficient to obtain radiation-resistant optical glass: the reduction in light transmission above 450 nm is nil at 10/sup 8/ rad (CH), below 10% at 10/sup 9/ rad (CH), and below 20% at 5*10/sup 9/ rad (CH); the post-irradiation fading is negligible.

Long-term behavior of nuclear glass: the r(t) operational model

International Nuclear Information System (INIS)

Ribet, I.; Gin, S.; Minet, Y.; Vernaz, E.; Chaix, P.; Do Quang, R.

2001-01-01

Predicting the long-term behavior of vitrified waste packages requires the development of models incorporating knowledge of the aqueous alteration mechanisms of nuclear glass. The r(t) model allows for the formation of a protective gel layer during leaching, and is thus able to account for the major drops in the glass alteration rate that are observed experimentally. This article describes the model hypotheses, the methodology implemented to determine its three internal parameters, and the results obtained from about fifty leaching experiments performed under various conditions. The orders of magnitude of the internal parameter values are indicated according to the alteration conditions. (author)

Solubility of metallic mercury in organic solvents; Solubilite du mercure metallique dans les solvants organiques; Rastvorimost' metallicheskoj rtuti v organicheskikh rastvoritelyakh; Solubilidad del mercurio metalico en solventes organicos

Energy Technology Data Exchange (ETDEWEB)

Klehr, E H; Voigt, A F [Institute for Atomic Research and Department of Chemistry, Iowa State University, Ames, IA (United States)

1962-03-15

The solubilities of metallic mercury at 25{sup o}C in the solvents carbon tetrachloride, 2, 2, 4-trimethylpentane, n-decane, benzene, toluene, nitrobenzene, chlorobenzene and bromobenzene have been determined by the use of Hg{sup 203} tracer. Measurements were made either by shaking mercury metal of known specific activity with the solvent, or by measuring the distribution coefficient of the metal between the solvent and water and combining this measurement with the solubility of mercury in water, or by both methods. Radioactivity was measured by counting precipitated samples with a GM counter or liquid aliquots with a well-type scintillator counter. All solubilities were in the range (4.6 to 16) x 10{sup -6} g atoms/l. For n-decane, toluene and chlorobenzene, the temperature dependence of the solubility was determined over the range 0 to 45{sup o}C. The Hildebrand-Scott theory of solubility has been applied to these systems for comparison. For the aliphatic hydrocarbons, the solubilities predicted by theory are within 35% of those observed, but for the other solvents, the predicted values are 4 to 18 times larger than those observed. Modification of the theory gives better agreement for some solvents but poorer agreement for others. Curves of log G (solubility) vs. 1/T were straight lines with slopes similar to those obtained from the Hildebrand-Scott theory but displaced from those curves. The slopes of the curves were used for the calculation of heats and entropies of solution. Entropies were also calculated from curves of log G vs. log T. The two sets of entropy values were in agreement with each other and were within one to four entropy units of the ideal entropies of mixing. (author) [French] Les auteurs ont determine la solubilite du mercure metallique a 25{sup o}C dans les solvants suivants: tetrachlorure de carbone, trimethylpentane 2, 2, 4, n-decane, benzene, toluene, nitrobenzene, chlorobenzene et bromobenzene en utilisant le mercure-203 comme

In situ one-year burial experiments with simulated nuclear waste glasses

International Nuclear Information System (INIS)

Hench, L.L.; Spilman, D.; Buonaquisti, T.; Werme, L.

1985-01-01

Two simulated nuclear waste glasses were corroded in an in-situ experiment in the Stripa mine up to one year at 90 degree C and ambient temperature. Changes in compositional in-depth profiles were measured using Fourier transform infrared reflection spectroscopy, SIMS and Rutherford back-scattering. For glass/glass interfaces, both glasses showed depletion of Na, Cs and B, but for the more corrosion resistant glass, the lower depletion is ascribed to the formation of a thin (0.2 nm) coherent and dense outer layer enriched in Mg, Ca, Sr, Ba, Zn-Al and Si, which impedes both ion exchange and network attack of the bulk underneath. For the bentonite interfaces, cation exchange of Ca, Mg, Al and Fe from the bentonite for primarily Na and B is found to produce a glass surface that has three silicate-rich layers. The larger concentrations of M/super2+/ and M/super3+/ cation and the high silica content of the reaction layers result in a considerably retarded rate of ion exchange after the formation of these layers during the first three months of burial. The granite interfaces showed the lowest rate of attack. This appears to be due to a large increase of Fe and Al within the glass surfaces exposed to granite. The results obtained using Rutheford back-scattering confirm the results obtained using the other techniques for surface analysis. Analysis of burial samples cast in steel mini-canisters show no significant effects associated with the steel canister-glass interface. (author)

FTIR spectra and properties of iron borophosphate glasses containing simulated nuclear wastes

Science.gov (United States)

Liao, Qilong; Wang, Fu; Chen, Kuiru; Pan, Sheqi; Zhu, Hanzhen; Lu, Mingwei; Qin, Jianfa

2015-07-01

30 wt.% simulated nuclear wastes were successfully immobilized by B2O3-doped iron phosphate base glasses. The structure and thermal stability of the prepared wasteforms were characterized by Fourier transform infrared spectroscopy and differential thermal analysis, respectively. The subtle structural variations attributed to different B2O3 doping modes have been discussed in detail. The results show that the thermal stability and glass forming tendency of the iron borophosphate glass wasteforms are faintly affected by different B2O3 doping modes. The main structural networks of iron borophosphate glass wasteforms are PO43-, P2O74-, [BO4] groups. Furthermore, for the wasteform prepared by using 10B2O3-36Fe2O3-54P2O5 as base glass, the distributions of Fe-O-P bonds, [BO4], PO43- and P2O74- groups are optimal. In general, the dissolution rate (DR) values of the studied iron borophosphate wasteforms are about 10-8 g cm-2 min-1. The obtained conclusions can offer some useful information for the disposal of high-level radioactive wastes using boron contained phosphate glasses.

Reuse of nuclear byproducts, NaF and HF in metal glass industries

Energy Technology Data Exchange (ETDEWEB)

Park, J.W.; Lee, H.W. [Korea Power Engineering Co., Inc., Kyunggi-do (Korea, Republic of); Yoo, S.H.; Moon, H.S.; Cho, N.C. [Korea Nuclear Fuel Co., Ltd., Daejon (Korea, Republic of)

1997-02-01

A study has been performed to evaluate the radiological safety and feasibility associated with reuse of NaF(Sodium Fluoride) and HF(Hydrofluoric Acid) which are generated as byproducts from the nuclear fuel fabrication process. The investigation of oversea`s experience reveals that the byproduct materials are most often used in the metal and glass industries. For the radiological safety evaluation, the uranium radioactivities in the byproduct materials were examined and shown to be less than radioactivities in natural materials. The radiation doses to plant personnel and the general public were assessed to be very small and could be ignored. The Korea nuclear regulatory body permits the reuse of NaF in the metal industry on the basis of associated radioactivity being {open_quote}below regulatory concern{close_quote}. HF is now under review for reuse acceptability in the steel and glass industries.

The effect of clay on the dissolution of nuclear waste glass

Science.gov (United States)

Lemmens, K.

2001-09-01

In a nuclear waste repository, the waste glass can interact with metals, backfill materials (if present) and natural host rock. Of the various host rocks considered, clays are often reported to delay the onset of the apparent glass saturation, where the glass dissolution rate becomes very small. This effect is ascribed to the sorption of silica or other glass components on the clay. This can have two consequences: (1) the decrease of the silica concentration in solution increases the driving force for further dissolution of glass silica, and (2) the transfer of relatively insoluble glass components (mainly silica) from the glass surface to the clay makes the alteration layer less protective. In recent literature, the latter explanation has gained credibility. The impact of the environmental materials on the glass surface layers is however not well understood. Although the glass dissolution can initially be enhanced by clay, there are arguments to assume that it will decrease to very low values after a long time. Whether this will indeed be the case, depends on the fate of the released glass components in the clay. If they are sorbed on specific sites, it is likely that saturation of the clay will occur. If however the released glass components are removed by precipitation (growth of pre-existing or new secondary phases), saturation of the clay is less likely, and the process can continue until exhaustion of one of the system components. There are indications that the latter mechanism can occur for varying glass compositions in Boom Clay and FoCa clay. If sorption or precipitation prevents the formation of protective surface layers, the glass dissolution can in principle proceed at a high rate. High silica concentrations are assumed to decrease the dissolution rate (by a solution saturation effect or by the impact on the properties of the glass alteration layer). In glass corrosion tests at high clay concentrations, silica concentrations are, however, often higher

The effect of clay on the dissolution of nuclear waste glass

International Nuclear Information System (INIS)

Lemmens, K.

2001-01-01

In a nuclear waste repository, the waste glass can interact with metals, backfill materials (if present) and natural host rock. Of the various host rocks considered, clays are often reported to delay the onset of the apparent glass saturation, where the glass dissolution rate becomes very small. This effect is ascribed to the sorption of silica or other glass components on the clay. This can have two consequences: (1) the decrease of the silica concentration in solution increases the driving force for further dissolution of glass silica, and (2) the transfer of relatively insoluble glass components (mainly silica) from the glass surface to the clay makes the alteration layer less protective. In recent literature, the latter explanation has gained credibility. The impact of the environmental materials on the glass surface layers is however not well understood. Although the glass dissolution can initially be enhanced by clay, there are arguments to assume that it will decrease to very low values after a long time. Whether this will indeed be the case, depends on the fate of the released glass components in the clay. If they are sorbed on specific sites, it is likely that saturation of the clay will occur. If however the released glass components are removed by precipitation (growth of pre-existing or new secondary phases), saturation of the clay is less likely, and the process can continue until exhaustion of one of the system components. There are indications that the latter mechanism can occur for varying glass compositions in Boom Clay and FoCa clay. If sorption or precipitation prevents the formation of protective surface layers, the glass dissolution can in principle proceed at a high rate. High silica concentrations are assumed to decrease the dissolution rate (by a solution saturation effect or by the impact on the properties of the glass alteration layer). In glass corrosion tests at high clay concentrations, silica concentrations are, however, often higher

Understanding of the mechanical and structural changes induced by alpha particles and heavy ions in the French simulated nuclear waste glass

International Nuclear Information System (INIS)

Karakurt, G.; Abdelouas, A.; Guin, J.-P.; Nivard, M.; Sauvage, T.; Paris, M.; Bardeau, J.-F.

2016-01-01

Borosilicate glasses are considered for the long-term confinement of high-level nuclear wastes. External irradiations with 1 MeV He + ions and 7 MeV Au 5+ ions were performed to simulate effects produced by alpha particles and by recoil nuclei in the simulated SON68 nuclear waste glass. To better understand the structural modifications, irradiations were also carried out on a 6-oxides borosilicate glass, a simplified version of the SON68 glass (ISG glass). The mechanical and macroscopic properties of the glasses were studied as function of the deposited electronic and nuclear energies. Alpha particles and gold ions induced a volume change up to −0.7% and −2.7%, respectively, depending on the glass composition. Nano-indentations tests were used to determine the mechanical properties of the irradiated glasses. A decrease of about −22% to −38% of the hardness and a decrease of the reduced Young's modulus by −8% were measured after irradiations. The evolution of the glass structure was studied by Raman spectroscopy, and also 11 B and 27 Al Nuclear Magnetic Resonance (MAS-NMR) on a 20 MeV Kr irradiated ISG glass powder. A decrease of the silica network connectivity after irradiation with alpha particles and gold ions is deduced from the structural changes observations. NMR spectra revealed a partial conversion of BO 4 to BO 3 units but also a formation of AlO 5 and AlO 6 species after irradiation with Kr ions. The relationships between the mechanical and structural changes are also discussed. - Highlights: • Mechanical and structural properties of two borosilicate glass compositions irradiated with alpha particles and heavy ions were investigated. • Both kinds of particles induced a decrease of the hardness, reduced Young's modulus and density. • Electronic and nuclear interactions are responsible for the changes observed. • The evolution of the mechanical properties under irradiation is linked to the changes occured in the

The use of positrons to survey alteration layers on synthetic nuclear waste glasses

International Nuclear Information System (INIS)

Reiser, Joelle T.; Parruzot, Benjamin; Weber, Marc H.; Ryan, Joseph V.; McCloy, John S.; Wall, Nathalie A.

2017-01-01

In order to safeguard society and the environment, understanding radioactive waste glass alteration mechanisms in interactions with solutions and near-field materials, such as Fe, is essential to nuclear waste repository performance assessments. Alteration products are formed at the surface of glasses after reaction with solution. In this study, glass altered in the presence of Fe 0 in aqueous solution formed two alteration layers: one embedded with Fe closer to the surface and one without Fe found deeper in the sample. Both layers were found to be thinner than the alteration layer found in glass altered in aqueous solution only. For the first time, Doppler Broadening Positron Annihilation Spectroscopy (DB-PAS) is used to non-destructively characterize the pore structures of glass altered in the presence of Fe 0 . Advantages and disadvantages of DB-PAS compared to other techniques used to analyze pore structures for altered glass samples are discussed. Ultimately, DB-PAS has shown to be an excellent choice for pore structure characterization for glasses with multiple alteration layers. Monte Carlo modeling predicted positron trajectories through the layers, and helped explain DB-PAS data, which showed that the deeper alteration layer without Fe had a similar composition and pore structure to layers on glass altered in water only.

The use of positrons to survey alteration layers on synthetic nuclear waste glasses

Energy Technology Data Exchange (ETDEWEB)

Reiser, Joelle T. [Washington State University, Chemistry Department, Pullman, WA 99164 (United States); Pacific Northwest National Laboratory, Energy and Environment Directorate, Richland, WA 99352 (United States); Parruzot, Benjamin [Pacific Northwest National Laboratory, Energy and Environment Directorate, Richland, WA 99352 (United States); Weber, Marc H. [Washington State University, Center for Materials Research, Pullman, WA 99164 (United States); Ryan, Joseph V., E-mail: joe.ryan@pnnl.gov [Pacific Northwest National Laboratory, Energy and Environment Directorate, Richland, WA 99352 (United States); McCloy, John S. [Washington State University, Chemistry Department, Pullman, WA 99164 (United States); Pacific Northwest National Laboratory, Energy and Environment Directorate, Richland, WA 99352 (United States); Washington State University, School of Mechanical and Materials Engineering, Pullman, WA 99164 (United States); Wall, Nathalie A., E-mail: nawall@wsu.edu [Washington State University, Chemistry Department, Pullman, WA 99164 (United States)

2017-07-15

In order to safeguard society and the environment, understanding radioactive waste glass alteration mechanisms in interactions with solutions and near-field materials, such as Fe, is essential to nuclear waste repository performance assessments. Alteration products are formed at the surface of glasses after reaction with solution. In this study, glass altered in the presence of Fe{sup 0} in aqueous solution formed two alteration layers: one embedded with Fe closer to the surface and one without Fe found deeper in the sample. Both layers were found to be thinner than the alteration layer found in glass altered in aqueous solution only. For the first time, Doppler Broadening Positron Annihilation Spectroscopy (DB-PAS) is used to non-destructively characterize the pore structures of glass altered in the presence of Fe{sup 0}. Advantages and disadvantages of DB-PAS compared to other techniques used to analyze pore structures for altered glass samples are discussed. Ultimately, DB-PAS has shown to be an excellent choice for pore structure characterization for glasses with multiple alteration layers. Monte Carlo modeling predicted positron trajectories through the layers, and helped explain DB-PAS data, which showed that the deeper alteration layer without Fe had a similar composition and pore structure to layers on glass altered in water only.

The use of positrons to survey alteration layers on synthetic nuclear waste glasses

Science.gov (United States)

Reiser, Joelle T.; Parruzot, Benjamin; Weber, Marc H.; Ryan, Joseph V.; McCloy, John S.; Wall, Nathalie A.

2017-07-01

In order to safeguard society and the environment, understanding radioactive waste glass alteration mechanisms in interactions with solutions and near-field materials, such as Fe, is essential to nuclear waste repository performance assessments. Alteration products are formed at the surface of glasses after reaction with solution. In this study, glass altered in the presence of Fe0 in aqueous solution formed two alteration layers: one embedded with Fe closer to the surface and one without Fe found deeper in the sample. Both layers were found to be thinner than the alteration layer found in glass altered in aqueous solution only. For the first time, Doppler Broadening Positron Annihilation Spectroscopy (DB-PAS) is used to non-destructively characterize the pore structures of glass altered in the presence of Fe0. Advantages and disadvantages of DB-PAS compared to other techniques used to analyze pore structures for altered glass samples are discussed. Ultimately, DB-PAS has shown to be an excellent choice for pore structure characterization for glasses with multiple alteration layers. Monte Carlo modeling predicted positron trajectories through the layers, and helped explain DB-PAS data, which showed that the deeper alteration layer without Fe had a similar composition and pore structure to layers on glass altered in water only.

SON68 nuclear glass dissolution kinetics: Current state of knowledge and basis of the new GRAAL model

Energy Technology Data Exchange (ETDEWEB)

Frugier, P. [CEA Marcoule, DTCD/SECM/LCLT, BP 17171, 30207 Bagnols-sur-Ceze cedex (France)], E-mail: pierre.frugier@cea.fr; Gin, S.; Minet, Y.; Chave, T. [CEA Marcoule, DTCD/SECM/LCLT, BP 17171, 30207 Bagnols-sur-Ceze cedex (France); Bonin, B. [CEA Saclay, DEN/DIR/DS, 91191 Gif-sur-Yvette cedex (France); Godon, N.; Lartigue, J.-E.; Jollivet, P. [CEA Marcoule, DTCD/SECM/LCLT, BP 17171, 30207 Bagnols-sur-Ceze cedex (France); Ayral, A. [IEM/CNRS-ENSCM Universite Montpellier 2, CC 047, Place Eugene Bataillon, 34095 Montpellier cedex 5 (France); De Windt, L. [ENSMP, CG, 35 rue St Honore, 77305 Fontainebleau cedex (France); Santarini, G. [CEA Saclay HC/CAB, 91191 Gif-sur-Yvette cedex (France)

2008-10-15

This article summarizes the present state of knowledge concerning aqueous alteration of R7T7-type nuclear containment glasses, represented mainly by the inactive reference glass designated SON68. Based on this review, we propose to describe the glass alteration kinetics up to and including the final residual rate regime by means of a new mechanistic model known as GRAAL (glassreactivitywithallowanceforthealterationlayer). Phenomenological analysis findings are reviewed for the various glass alteration regimes: interdiffusion, initial rate, rate drop, residual rate and, under very particular circumstances, resumption of alteration. These alteration regimes are associated with predominant mechanisms. Published work interpreting and modeling these mechanisms was examined in detail. There is a broad consensus on the general mechanisms of the initial rate and even the interdiffusion regime, whereas the mechanisms controlling the rate drop remain a subject of dispute not only with regard to nuclear glasses but also for the dissolution of silicate minerals. The reaction affinity responsible for the rate drop is expressed differently by different authors and depending on the underlying theories. The disagreement concerns the nature of the phase (glass or gel) or the activated complex controlling the rate drop, which in turn determines the elements that must be taken into account in the overall affinity term. Progress in recent years, especially in identifying the mechanisms responsible for the residual rate, has shed new light on these issues, allowing us to propose new theoretical foundations for modeling the different kinetic regimes of SON68 nuclear glass dissolution. The GRAAL model considers that water diffusion in the passivating reaction zone (the gel formed under saturation conditions) is a rate-limiting step in the overall glass dissolution kinetics. Moreover, this passivation zone is a soluble phase whose stability is directly dependent on the nature of the

Lead-iron phosophate glass

International Nuclear Information System (INIS)

Sales, B.C.; Boatner, L.A.

1988-01-01

The lead-iron phosphate nuclear waste glasses (LIPNWG) are the subject of the present chapter. They were discovered in 1984 while the authors were attempting to find a sintering aid for certain types of crystalline monazite ceramic high-level nuclear waste forms. In the present chapter, the term waste glass is synonymous with nuclear waste glass (NWG), and the acronym LIP is often used for lead-iron phosphate. Lead-iron phosphate glasses, like many of the previously studied phosphate glasses, are corrosion resistant in aqueous solutions at temperatures below 100 degrees C, and they can be melted and poured at temperatures that are relatively low in comparison with the processing temperatures required for current silicate glass compositions. Unlike the phosphate glasses investigated previously, however, LIPNWGs do not suffer from alteration due to devitrification during realistic and readily, achievable cooling periods. Additionally, lead-iron phosphate glass melts are not nearly as corrosive as the sodium phosphate melts investigated during the 1960s; and, therefore, they can be melted and processed using crucibles made from a variety of materials

Tests for determining impact resistance and strength of glass used for nuclear waste disposal

International Nuclear Information System (INIS)

Bunnell, L.R.

1979-05-01

Tests are described for determining the impact resistance (Section A) and static tensile strength (Section B) of glasses containing simulated or actual nuclear wastes. This report describes the development and use of these tests to rank different glasses, to assess effects of devitrification, and to examine the effect of impact energy on resulting surface area. For clarity this report is divided into two sections, Impact Resistance and Tensile Strength

Quantification of the Partitioning Ratio of Minor Actinide Surrogates between Zirconolite and Glass in Glass-Ceramic for Nuclear Waste Disposal.

Science.gov (United States)

Liao, Chang-Zhong; Liu, Chengshuai; Su, Minhua; Shih, Kaimin

2017-08-21

Zirconolite-based glass-ceramic is considered a promising wasteform for conditioning minor actinide-rich nuclear wastes. Recent studies on this wasteform have sought to enhance the partitioning ratio (PR) of minor actinides in zirconolite crystal. To optimize the PR in the SiO 2 -Al 2 O 3 -CaO-TiO 2 -ZrO 2 system, a novel conceptual approach, which can be derived from the chemical composition and quantity of zirconolite crystal in glass-ceramic, was introduced based on the results of Rietveld quantitative X-ray diffraction analysis and transmission electron microscopy energy dispersive X-ray spectroscopy. To verify this new conceptual approach, the influences of the crystallization temperature, the concentration of additives, and ionic radii on the PR of various surrogates (Ce, Nd, Gd, and Yb) in zirconolite were examined. The results reveal that the PR of Nd 3+ in zirconolite can be as high as 41%, but it decreases as the crystallization temperature increases. The quantities of all phases (including crystalline and amorphous) remained nearly constant when increasing the loading of Nd 2 O 3 in glass-ceramic products crystallized at 1050 °C for 2 h. Correspondingly, the PR of Nd 3+ decreases in a linear fashion with the loading contents of Nd 2 O 3 . The radius of ions also has a great influence on the PR, and an increase in the ionic radius leads to a decrease in the PR. This new approach will be an important tool to facilitate the exploration of a glass-ceramic matrix for the disposal of minor actinide-rich nuclear wastes.

Comprehension and modelling of chromia-forming alloys corrosion mechanisms in nuclear glasses

International Nuclear Information System (INIS)

Schmucker, Eric

2016-01-01

Nuclear wastes management consists in the confinement of the radioactive wastes in a glass matrix. This is made by inductive melting in a hot crucible at an operating temperature around 1150 C. These crucibles are constituted of nickel based superalloys with high chromium content. They are submitted to a harsh corrosion by the molten glass, eventually leading to their replacement. The protection of the crucible against corrosion is best provided by the establishment of a protective chromium oxide layer at the surface of the alloy. A binary chromia-forming alloy (Ni-30Cr) is studied in this work. Three different binary and ternary glass compositions are chosen in order to understand the influence of the glass basicity and glass viscosity on the corrosion kinetics. Besides, the de-correlation of the formation and dissolution kinetics of the oxide layer allows the modelling of the overall oxide growth in the molten glass. For that purpose, the oxide formation kinetics in molten glass media is assimilated to the oxidation kinetics of the alloy in gaseous media with oxygen partial pressure that are representative of the redox properties of the glasses. Studies of the oxidation kinetics and of the diffusion mechanisms have shown that the oxidation kinetics is independent on the oxygen pressure in the range of 10"-"1"3 up to 10"-"3 atm O_2 at 1150 C. The present work has shown that the dissolution kinetics of the oxide layer is governed by the diffusion of Cr(III) in the glass melt. This dissolution kinetics has been evaluated from the diffusion coefficient and the solubility limit of Cr(III) in the glass. Finally, the overall growth kinetics of the Cr_2O_3 layer in the glass has been successfully modelled for each glass, thanks to the knowledge of (i) the solubility limit of Cr(III), (ii) its diffusion coefficient in the glasses and (iii) the oxidation kinetics of the alloy. The presented model also allows quantifying the influence of each of these parameters on the

Understanding of the mechanical and structural changes induced by alpha particles and heavy ions in the French simulated nuclear waste glass

Energy Technology Data Exchange (ETDEWEB)

Karakurt, G., E-mail: karakurt_gokhan@yahoo.fr [SUBATECH, UMR 6457CNRS-IN2P3, Ecole des Mines de Nantes, 4 rue Alfred Kastler, 44307 Nantes (France); Abdelouas, A. [SUBATECH, UMR 6457CNRS-IN2P3, Ecole des Mines de Nantes, 4 rue Alfred Kastler, 44307 Nantes (France); Guin, J.-P.; Nivard, M. [Institut de Physique de Rennes, Université de Rennes 1 – UMR 62051 IPR, 263 avenue du Général Leclerc, 35042 Rennes (France); Sauvage, T. [Laboratoire CEMHTI (Conditions Extrêmes et Matériaux: Haute Température et Irradiation), CNRS UPR, 3079 Orléans (France); Paris, M. [Institut des Matériaux Jean ROUXEL, Université de Nantes, UMR 6502 CNRS, 2 rue de la Houssinière, BP 32229, 44322 Nantes Cedex 03 (France); Bardeau, J.-F. [Institut des Molécules et Matériaux du Mans, UMR CNRS 6283, avenue Olivier Messiaen, 72085 Le Mans (France)

2016-07-15

Borosilicate glasses are considered for the long-term confinement of high-level nuclear wastes. External irradiations with 1 MeV He{sup +} ions and 7 MeV Au{sup 5+} ions were performed to simulate effects produced by alpha particles and by recoil nuclei in the simulated SON68 nuclear waste glass. To better understand the structural modifications, irradiations were also carried out on a 6-oxides borosilicate glass, a simplified version of the SON68 glass (ISG glass). The mechanical and macroscopic properties of the glasses were studied as function of the deposited electronic and nuclear energies. Alpha particles and gold ions induced a volume change up to −0.7% and −2.7%, respectively, depending on the glass composition. Nano-indentations tests were used to determine the mechanical properties of the irradiated glasses. A decrease of about −22% to −38% of the hardness and a decrease of the reduced Young's modulus by −8% were measured after irradiations. The evolution of the glass structure was studied by Raman spectroscopy, and also {sup 11}B and {sup 27}Al Nuclear Magnetic Resonance (MAS-NMR) on a 20 MeV Kr irradiated ISG glass powder. A decrease of the silica network connectivity after irradiation with alpha particles and gold ions is deduced from the structural changes observations. NMR spectra revealed a partial conversion of BO{sub 4} to BO{sub 3} units but also a formation of AlO{sub 5} and AlO{sub 6} species after irradiation with Kr ions. The relationships between the mechanical and structural changes are also discussed. - Highlights: • Mechanical and structural properties of two borosilicate glass compositions irradiated with alpha particles and heavy ions were investigated. • Both kinds of particles induced a decrease of the hardness, reduced Young's modulus and density. • Electronic and nuclear interactions are responsible for the changes observed. • The evolution of the mechanical properties under irradiation is linked

Glass science tutorial: Lecture No. 7, Waste glass technology for Hanford

International Nuclear Information System (INIS)

Kruger, A.A.

1995-07-01

This paper presents the details of the waste glass tutorial session that was held to promote knowledge of waste glass technology and how this can be used at the Hanford Reservation. Topics discussed include: glass properties; statistical approach to glass development; processing properties of nuclear waste glass; glass composition and the effects of composition on durability; model comparisons of free energy of hydration; LLW glass structure; glass crystallization; amorphous phase separation; corrosion of refractories and electrodes in waste glass melters; and glass formulation for maximum waste loading

Effects of alpha decays on nuclear waste glasses, simulation through atomistic models

International Nuclear Information System (INIS)

Ghaleb, D.; Delaye, J.M.

1997-01-01

In a simplified (SiO 2 , B 2 O 3 , Na 2 O 3 , Al 2 O 3 , ZrO 2 ) nuclear glass we have simulated, by Molecular Dynamics simulations, the effects of displacement cascades created by the slowing-down of the recoil nucleus. The methodology employed to construct and validate the used Molecular Dynamics model representing the basis matrix of the 'light-water' French nuclear glass (R77) and the manner which are simulated atomic displacements are described. Although the energies given to recoil nucleus were relatively low (≤ 1/10 of actual energies) the study has yielded a number of interesting results. Notably we have: - identified the main mechanisms responsible for the depolymerization of the network; - observed, at the atomic level, the kinetic of the structure evolution; - detailed the behavior and displacement mechanisms of every atomic species during the cascade sequences; - made a link with the experimentation through the calculation of some physical properties. (authors)

Raman and X-ray absorption spectroscopic studies of hydrothermally altered alkali-borosilicate nuclear waste glass

Energy Technology Data Exchange (ETDEWEB)

McKeown, David A., E-mail: davidm@vsl.cua.ed [Vitreous State Laboratory, Catholic University of America, 620 Michigan Ave., N.E., Washington, DC 20064 (United States); Buechele, Andrew C.; Viragh, Carol; Pegg, Ian L. [Vitreous State Laboratory, Catholic University of America, 620 Michigan Ave., N.E., Washington, DC 20064 (United States)

2010-04-01

Raman spectroscopy and X-ray absorption spectroscopy (XAS) are used to characterize structural changes that took place in hydrothermally altered (Na,K)-alumina-borosilicate glasses with different Na/K ratios, formulated as part of a durability study to investigate the behavior of glasses for nuclear waste storage. The hydrothermal experiments, or vapor hydration tests (VHT), were performed on each glass for 3 and 20 days at 200 deg. C to accelerate and approximate long-term alteration processes that may occur in a nuclear waste repository. Results found for both glasses and their VHT altered counterparts show little, if any, structural influence from the different starting Na/K ratios. X-ray diffraction, differential scanning calorimetry, scanning electron microscopy, and Raman spectroscopy indicate that the altered samples are mostly amorphous with small amounts of analcime-like and leucite-like crystals within 200 mum of the sample surface and contain up to 9.7 wt.% water or OH. The Raman data are nearly identical for the amorphous portions of all altered VHT samples investigated, and indicate that two glass structural changes took place during alteration: one, partial depolymerization of the alumina-borosilicate network, and two, introduction of water or OH. Al and Si XAS data indicate tetrahedral AlO{sub 4} and SiO{sub 4} environments in the original glasses as well as in the altered samples. Small energy shifts of the Si K-edge also show that the altered VHT samples have less polymerized networks than the original glass. Na XAS data indicate expanded Na environments in the VHT samples with longer Na-O distances and more nearest-neighbor oxygen atoms, compared with the original glasses, which may be due to hydrous species introduced into the expanding Na-sites.

Understanding the origin of the fracture toughness evolution of nuclear glasses under irradiation

International Nuclear Information System (INIS)

Kieu, L.-H.

2011-01-01

In the nuclear industry, complex borosilicate glasses are used for the confinement of fission products and long-life minor actinides. Under irradiations, the structure and the mechanical properties of these glasses evolve. In this work, atomistic and multi-scale simulations of three simplified borosilicate glasses were run to understand the origin of their fracture behavior change under irradiation. Under the radiation effects, elasticity decreases and plasticity increases. Fracture happens due to the formation and coalescence of nano-cavities. The structural modifications under the radiation effects lead to a delay of the coalescence and of the irradiated glass rupture. Several phenomena overlay to explain this behavior, especially the cavities distribution modifications, the sodium mobility, and the borate and silicate entities organization in the glassy network. Depending on the nature of the more important mechanism, the fracture toughness can increase or decrease under radiation. (author) [fr

Porous glass with high silica content for nuclear waste storage : preparation, characterization and leaching

International Nuclear Information System (INIS)

Aegerter, M.A.; Santos, D.I. dos; Ventura, P.C.S.

1984-01-01

Aqueous solutions simulating radioactive nuclear wastes (like Savanah River Laboratory) were incorporated in porous glass matrix with high silica content prepared by decomposition of borosilicate glass like Na 2 O - B 2 O 3 - SiO 2 . After sintering, the samples were submitted, during 28 days, to standard leaching tests MCC1, MCC5 (Soxhlet) and stagnating. The total weight loss, ph, as well as the integral and differential leaching rates and the accumulated concentrations in the leach of Si, Na, B, Ca, Mn, Al, Fe and Ni. The results are compared with the results from reference borosilicate glass, made by fusion, ceramic, synroc, concrets, etc... (E.G.) [pt

Electrical power supply and controls for a remotely operated glass melter for nuclear waste

International Nuclear Information System (INIS)

Haideri, A.Q.

1985-01-01

An electrical power supply, controls and instruments used for a joule heated glass melter for nuclear waste are discussed. Remotely replaceable interconnection wiring assemblies for power, controls and instruments are also described

Glass microspheres for brachytherapy

International Nuclear Information System (INIS)

Prado, Miguel O.; Prastalo, Simon; Blaumann, Herman; Longhino, Juan M.; Repetto Llamazares, A.H.V.

2007-01-01

We developed the capacity to produce glass microspheres containing in their structure one or more radioactive isotopes useful for brachytherapy. We studied the various facts related with their production: (Rare earth) alumino silicate glass making, glass characterization, microspheres production, nuclear activation through (n,γ) nuclear reactions, mechanical characterization before and after irradiation. Corrosion tests in simulated human plasma and mechanical properties characterization were done before and after irradiation. (author) [es

3 and 4 oxidation state element solubilities in borosilicate glasses. Implement to actinides in nuclear glasses

International Nuclear Information System (INIS)

Cachia, J.N.

2005-12-01

In order to ensure optimal radionuclides containment, the knowledge of the actinide loading limits in nuclear waste glasses and also the comprehension of the solubilization mechanisms of these elements are essential. A first part of this manuscript deals with the study of the differences in solubility of the tri and tetravalent elements (actinides and surrogates) particularly in function of the melting temperature. The results obtained indicate that trivalent elements (La, Gd, Nd, Am, Cm) exhibit a higher solubility than tetravalent elements (Hf, Th, Pu). Consequently, it was planned to reduce plutonium at the oxidation state (III), the later being essentially tetravalent in borosilicate glasses. An innovating reduction process of multi-valent elements (cerium, plutonium) using silicon nitride has been developed in a second part of this work. Reduced plutonium-bearing glasses synthesized by Si 3 N 4 addition made it possible to double the plutonium solubility from 2 to 4 wt% at 1200 deg C. A structural approach to investigate the differences between tri and tetravalent elements was finally undertaken. These investigations were carried out by X-rays Absorption Spectroscopy (EXAFS) and NMR. Trivalent rare earth and actinide elements seem to behave as network modifiers while tetravalent elements rather present true intermediaries' behaviour. (author)

MASBAL: A computer program for predicting the composition of nuclear waste glass produced by a slurry-fed ceramic melter

International Nuclear Information System (INIS)

Reimus, P.W.

1987-07-01

This report is a user's manual for the MASBAL computer program. MASBAL's objectives are to predict the composition of nuclear waste glass produced by a slurry-fed ceramic melter based on a knowledge of process conditions; to generate simulated data that can be used to estimate the uncertainty in the predicted glass composition as a function of process uncertainties; and to generate simulated data that can be used to provide a measure of the inherent variability in the glass composition as a function of the inherent variability in the feed composition. These three capabilities are important to nuclear waste glass producers because there are constraints on the range of compositions that can be processed in a ceramic melter and on the range of compositions that will be acceptable for disposal in a geologic repository. MASBAL was developed specifically to simulate the operation of the West Valley Component Test system, a commercial-scale ceramic melter system that will process high-level nuclear wastes currently stored in underground tanks at the site of the Western New York Nuclear Services Center (near West Valley, New York). The program is flexible enough, however, to simulate any slurry-fed ceramic melter system. 4 refs., 16 figs., 5 tabs

Cumulative and competitive effects of chemical elements on nuclear glass alteration

International Nuclear Information System (INIS)

Arena, Helene

2016-01-01

This work takes place in the context of the long-term behavior of nuclear glasses under repository conditions. The main objective is to identify, understand and compare the effects of some chemical elements present in the glass composition and/or in the repository media (Zn, Mg, Ni, Co, Fe, Ca, Gd, Ce, K, Cs, Cr and Ag) on the processes involved in glass alteration by water. The cumulative or competitive nature of the effects of these chemical elements was determined. To reach this goal, a 6 oxides simple glass (ISG) has been altered for more than 500 days in a solution containing one or more of the chemical elements of interest. The results indicate that Zn, Mg, Ni, Co and Fe elements increase glass alteration forming secondary phases with the same structure and stoichiometry (tri-octahedral smectites). To form, these silicates consume chemical elements (Si, Al) from the environment and induce a pH decrease until a limiting value of pH. Beyond this pH the precipitation of secondary phases is inhibited and these chemical elements can be integrated into the gel, replacing Ca whose solubility increases at lower pH. As long as they form secondary phases, the effects of these elements are cumulative. Rare earths Gd and Ce also increase glass alteration forming secondary phases but their effects are lower as they contain less silicon. These elements are not integrated in the gel. Chromium increases glass alteration by precipitating with Ca and leading to a less protective gel, depleted in Ca. Silver precipitates as AgCl and has no effect on the alteration of the glass. The chemical elements K, Cs and Ca limit glass alteration by integrating into the gel and slowing down the transport phenomena therein. This integration is competitive: the order of integration (quantity and effectiveness glass alteration limitation) is the following Ca≥≥Cs≥K. Thus, the increase of glass alteration may be proportional to the quantity of elements promoting the precipitation of

Effect of neutron irradiation on etching, optical and structural properties of microscopic glass slide used as a solid state nuclear track detector

International Nuclear Information System (INIS)

Singh, Surinder; Kaur Sandhu, Amanpreet; Prasher, Sangeeta; Prakash Pandey, Om

2007-01-01

Microscopic glass slides are soda-lime glasses which are readily available and are easy to manufacture with low production cost. The application of these glasses as nuclear track detector will help us to make use of these glasses as solid-state nuclear track detector. The present paper describes the variation in the etching, optical and structural properties of the soda-lime microscopic glass slides due to neutron irradiation of different fluences. The color transformation and an increase in the optical absorption with neutron irradiation are observed. Both the bulk and track etch rates are found to increase with neutron fluence, thus showing a similar dependence on neutron fluence, but the sensitivity remains almost constant

Review of glass ceramic waste forms

International Nuclear Information System (INIS)

Rusin, J.M.

1981-01-01

Glass ceramics are being considered for the immobilization of nuclear wastes to obtain a waste form with improved properties relative to glasses. Improved impact resistance, decreased thermal expansion, and increased leach resistance are possible. In addition to improved properties, the spontaneous devitrification exhibited in some waste-containing glasses can be avoided by the controlled crystallization after melting in the glass-ceramic process. The majority of the glass-ceramic development for nuclear wastes has been conducted at the Hahn-Meitner Institute (HMI) in Germany. Two of their products, a celsian-based (BaAl 3 Si 2 O 8 ) and a fresnoite-based (Ba 2 TiSi 2 O 8 ) glass ceramic, have been studied at Pacific Northwest Laboratory (PNL). A basalt-based glass ceramic primarily containing diopsidic augite (CaMgSi 2 O 6 ) has been developed at PNL. This glass ceramic is of interest since it would be in near equilibrium with a basalt repository. Studies at the Power Reactor and Nuclear Fuel Development Corporation (PNC) in Japan have favored a glass-ceramic product based upon diopside (CaMgSi 2 O 6 ). Compositions, processing conditions, and product characterization of typical commercial and nuclear waste glass ceramics are discussed. In general, glass-ceramic waste forms can offer improved strength and decreased thermal expansion. Due to typcially large residual glass phases of up to 50%, there may be little improvement in leach resistance

Study of nuclear glasses alteration gel and synthesis of some model gels

International Nuclear Information System (INIS)

Ricol, S.

1995-01-01

This work deals with the general problem of alteration of the reference nuclear glass R7T7. Attention is paid particularly to the altered layer formed at the glass surface during alteration process. In opposition to previous works, related essentially to glass dissolution kinetics based on chemical analyses of lixiviated elements, this thesis deals with alteration problems through structural studies of the reference glass and derived gel. This approach allows the determination of mechanisms for the gel formation and a better understanding of the behaviour of glasses towards lixiviation. Both approaches appeared complementary. Based on several spectroscopic techniques, this work showed the particular role of cations such as calcium, zirconium and iron. Studies of silica-based synthetic gels showed the synergic effect of formers cation and of one highly coordinated cation. The variation of the wavenumber related to Si-O-Si asymmetric stretching vibration can be correlated to O/Si ratio for ternary systems Si/Na/Zr. On the contrary, the Si losses of the materials depend on the number of non-bridging oxygen atoms. In the perspective of long-term behaviour, the alteration gel presents better characteristics than initial glass. It is therefore a highly stable material in static conditions. In the same way, synthetic gels are materials with very low solubilities (much lower than the alteration gel) and could be used as confining matrices. (authors). refs., 71 figs., 37 tabs

Glass corrosion in natural environments

Science.gov (United States)

Thorpe, Arthur N.; Barkatt, Aaron

1992-01-01

Experiments carried out during the progress period are summarized. Experiments carried out involving glass samples exposed to solutions of Tris have shown the appearance of 'spikes' upon monitoring glass dissolution as a function of time. The periodic 'spikes' observed in Tris-based media were interpreted in terms of cracking due to excessive stress in the surface region of the glass. Studies of the interactions of silicate glasses with metal ions in buffered media were extended to systems containing Al. Caps buffer was used to establish the pH. The procedures used are described and the results are given. Preliminary studies were initiated as to the feasibility of adding a slowly dissolving solid compound of the additive to the glass-water system to maintain a supply of dissolved additive. It appears that several magnesium compounds have a suitable combination of solubility and affinity towards silicate glass surfaces to have a pronounced retarding effect on the extraction of uranium from the glass. These preliminary findings raise the possibility that introducing a magnesium source into geologic repositories for nuclear waste glass in the form of a sparingly soluble Mg-based backfill material may cause a substantial reduction in the extent of long-term glass corrosion. The studies described also provide mechanistic understanding of the roles of various metal solutes in the leachant. Such understanding forms the basis for developing long-term predictions of nuclear waste glass durability under repository conditions. From what is known about natural highly reduced glasses such as tektites, it is clear that iron is dissolved as ferrous iron with little or no ferric iron. The reducing conditions were high enough to cause metallic iron to exsolve out of the glass in the form of submicroscopic spherules. As the nuclear waste glass is much less reduced, a study was initiated on other natural glasses in addition to the nuclear waste glass. Extensive measurements were

Compositional threshold for Nuclear Waste Glass Durability

International Nuclear Information System (INIS)

Kruger, Albert A.; Farooqi, Rahmatullah; Hrma, Pavel R.

2013-01-01

Within the composition space of glasses, a distinct threshold appears to exist that separates 'good' glasses, i.e., those which are sufficiently durable, from 'bad' glasses of a low durability. The objective of our research is to clarify the origin of this threshold by exploring the relationship between glass composition, glass structure and chemical durability around the threshold region

The development of basic glass formulations for solidifying HLW from nuclear fuel reprocessing plant

International Nuclear Information System (INIS)

Jiang Yaozhong; Tang Baolong; Zhang Baoshan; Zhou Hui

1995-01-01

Basic glass formulations 90U/19, 90U/20, 90Nd/7 and 90Nd/10 applied in electric melting process are developed by using the mathematical model of the viscosity and electric resistance of waste glass. The yellow phase does not occur for basic glass formulations 90U/19 and 90U/20 solidifying HLW from nuclear fuel reprocessing plant when the waste loading is 20%. Under the waste loading is 16%, the process and product properties of glass 90U/19 and 90U/20 come up to or surpass the properties of the same kind of foreign waste glasses, and other properties are about the same to them of foreign waste glasses. The process and product properties of basic glass formulations 90Nd/7 and 90Nd/10 used for the solidification of 'U replaced by Nd' liquid waste are almost similar to them of 90U/19 and 90U/20. These properties fairly meet the requirements of 'joint test' (performed at KfK-INE, Germany). Among these formulations, 90Nd/7 is applied in cold engineering scale electric melting test performed at KfK-INE in Germany. The main process properties of cold test is similar to laboratory results

Raman and 11B nuclear magnetic resonance spectroscopic studies of alkaline-earth lanthanoborate glasses

International Nuclear Information System (INIS)

Brow, R.K.; Tallant, D.R.; Turner, G.L.

1996-01-01

Glasses from the RO·La 2 O 3 ·B 2 O 3 (R = Mg, Ca, and Ba) systems have been examined. Glass formation is centered along the metaborate tie line, from La(BO 2 ) 3 to R(BO 2 ) 2 . Glasses generally have transition temperatures >600 C and expansion coefficients between 60 x 10 -7 /C and 100 x 10 -7 /C. Raman and solid-state nuclear magnetic resonance spectroscopies reveal changes in the metaborate network that depend on both the [R]:[La] ratio and the type of alkaline-earth ion. The fraction of tetrahedral sites is generally reduced in alkaline-earth-rich glasses, with magnesium glasses possessing the lowest concentration of B[4]. Raman spectra indicate that, with increasing [R]:[La] ratio, the preferred metaborate anion changes from a double-chain structure associated with crystalline La(BO 2 ) 3 to the single-chain and ring metaborate anions found in crystalline R(BO 2 ) 2 phases. In addition, disproportionation of the metaborate anions leads to the formation of a variety of other species, including pyroborates with terminal oxygens and more-polymerized species, such as diborates, with tetrahedral borons. Such structural changes are related to the ease of glass formation and some of the glass properties

Glass fiber sensors for detecting special nuclear materials at portal and monitor stations

International Nuclear Information System (INIS)

Hull, C.D.; Seymour, R.; Crawford, T.; Bliss, M.; Craig, R.A.

2001-01-01

Nuclear Safeguards and Security Systems LLC (NucSafe) participated in the Illicit Trafficking Radiation Assessment Program (ITRAP) recently conducted by the Austrian Research Center, Seibersdorf (ARCS) for IAEA, INTERPOL, and the World Customs Organization (IAEA, in press). This presentation reviews ITRAP test results of NucSafe instrumentation. NucSafe produces stationary, mobile, and hand-held systems that use neutron and gamma ray sensors to detect Special Nuclear Materials (SNM). Neutron sensors are comprised of scintillating glass fibers (trade name 'PUMA' for Pu Materials Analysis), which provide several advantages over 3 He and 10 BF 3 tubes. PUMA 6 Li glass fiber sensors offer greater neutron sensitivity and dynamic counting range with significantly less microphonic susceptibility than tubes, while eliminating transport and operational hazards. PUMA sensors also cost less per active area than gas tubes, which is important since rapid neutron detection at passenger, freight, and vehicle portals require large sensor areas to provide the required sensitivity

The influence of penetrating gamma radiation on the reaction of simulated nuclear waste glass in tuff groundwater

International Nuclear Information System (INIS)

Ebert, W.L.; Bates, J.K.; Abrajano, T.A. Jr.; Gerding, T.J.

1989-01-01

Static leaching experiments have been performed to determine the influence of penetrating gamma radiation on the reaction of simulated nuclear waste glass in tuff groundwater at 90/degree/C. Both the leachates and the reacted glass monoliths were analyzed to characterize the reaction. Radiation was seen to acidify the leachates, but the high bicarbonate content of the groundwater prevented the pHs from dropping below 6.4. The glass reaction tended to raise the pH. Glass based on SRL 165 black frit and PNL 76-68 glass compositions were leached. The SRL 165 type glasses were quite durable and unaffected by radiation [NL(B) /approximately/4 g/m 2 after 278 days at all exposure rates]. The PNL 76-68 glasses were much less durable, with the durability decreasing as the exposure rate was increased [NL(B) was about 20 g/m 2 after 278 days at 1 /times/ 10 3 R/h for both ATM-1c and ATM-8 glasses]. 8 refs., 5 figs

Silicate glasses. Chapter 1

International Nuclear Information System (INIS)

Lutze, W.

1988-01-01

This chapter is a survey of world-wide research and development efforts in nuclear waste glasses and its production technology. The principal glasses considered are silicate glasses which contain boron, i.e. borosilicate glass. A historical overview of waste form development programs in nine countries is followed by a summary of the design criteria for borosilicate glass compositions glass compositions. In the sections on glass properties the waste form is characterized in terms of potential alterations under the influence of heat, thermal gradients, radiation, aqueous solutions and combinations thereof. The topics are phase transformations, mechanical properties, radiation effects and chemical durability. The results from studies of volcanic glasses, as natural analogues for borosilicate nuclear waste glasses in order to verify predictions obtained from short-term tests in the laboratory, have been compiled in a special section on natural analogues. A special section on advanced vitrification techniques summarizes the various actual and potential processing schemes and describes the facilities. The literature has been considered until 1985. (author). 430 refs.; 68 figs.; 29 tabs

Statistical evaluation of the analytical method involved in French nuclear glasses leaching rate determination

Energy Technology Data Exchange (ETDEWEB)

Broudic, V.; Marques, C.; Bonnal, M

2004-07-01

Chemical durability studies of nuclear glasses involves a large number of water leaching experiments at different temperatures and pressures on both, glasses doped with fission products and actinides and non radioactive surrogates. The leaching rates of these glasses are evaluated through ICPAES analysis of the leachate over time. This work presents a statistical evaluation of the analysis method used to determine the concentrations of various vitreous matrix constituents: Si, B, Na, Al, Ca, Li as major elements and Ba, Cr, Fe, Mn, Mo, Ni, P, Sr, Zn, Zr as minor elements. Calibration characteristics, limits of detection, limits of quantification and uncertainties quantification are illustrated with different examples of analysis performed on surrogates and on radioactive leachates in glove box. (authors)

Statistical evaluation of the analytical method involved in French nuclear glasses leaching rate determination

International Nuclear Information System (INIS)

Broudic, V.; Marques, C.; Bonnal, M.

2004-01-01

Chemical durability studies of nuclear glasses involves a large number of water leaching experiments at different temperatures and pressures on both, glasses doped with fission products and actinides and non radioactive surrogates. The leaching rates of these glasses are evaluated through ICPAES analysis of the leachate over time. This work presents a statistical evaluation of the analysis method used to determine the concentrations of various vitreous matrix constituents: Si, B, Na, Al, Ca, Li as major elements and Ba, Cr, Fe, Mn, Mo, Ni, P, Sr, Zn, Zr as minor elements. Calibration characteristics, limits of detection, limits of quantification and uncertainties quantification are illustrated with different examples of analysis performed on surrogates and on radioactive leachates in glove box. (authors)

Structure study and properties of rare earth-rich glassed for the conditioning of nuclear waste

International Nuclear Information System (INIS)

Bardez, I.

2004-11-01

A new nuclear glass composition, able to immobilize highly radioactive liquid wastes from high burn-up UO 2 fuel, was established and its structure studied. The composition of the selected rare earth-rich glass is (molar %): 61.79 SiO 2 - 8.94 B 2 O 3 - 3.05 Al 2 O 3 - 14.41 Na 2 O - 6.32 CaO - 1.89 ZrO 2 - 3.60 RE 2 O 3 (with RE = La, Ce, Pr and Nd) The aim of this study was to determine the local environment of the rare earth in this glass and also to glean information about the effect of glass composition on the rare earth neighbouring (influence of Si, B, Al, Na and Ca contents). To this end, several series of glasses, prepared from the baseline glass, were studied by different characterisation methods such as EXAFS spectroscopy at the neodymium L III -edge, optical absorption spectroscopy, Raman spectroscopy and 29 Si, 27 Al and 11 B MAS-NMR. By coupling all the results obtained, several hypotheses about the nature of the rare earth neighbouring in the glass were proposed. (author)

Temperature effects on waste glass performance

International Nuclear Information System (INIS)

Mazer, J.J.

1991-02-01

The temperature dependence of glass durability, particularly that of nuclear waste glasses, is assessed by reviewing past studies. The reaction mechanism for glass dissolution in water is complex and involves multiple simultaneous reaction proceeded, including molecular water diffusion, ion exchange, surface reaction, and precipitation. These processes can change in relative importance or dominance with time or changes in temperature. The temperature dependence of each reaction process has been shown to follow an Arrhenius relationship in studies where the reaction process has been isolated, but the overall temperature dependence for nuclear waste glass reaction mechanisms is less well understood, Nuclear waste glass studies have often neglected to identify and characterize the reaction mechanism because of difficulties in performing microanalyses; thus, it is unclear if such results can be extrapolated to other temperatures or reaction times. Recent developments in analytical capabilities suggest that investigations of nuclear waste glass reactions with water can lead to better understandings of their reaction mechanisms and their temperature dependences. Until a better understanding of glass reaction mechanisms is available, caution should be exercised in using temperature as an accelerating parameter. 76 refs., 1 tab

Aqueous solutions/nuclear glasses interactions

International Nuclear Information System (INIS)

Delage, F.; Advocat, T.; Vernaz, E.; Crovisier, J.L.

1991-01-01

Interactions results of the borosilicate glass used in radioactive wastes confinement and aqueous solutions at various temperature and PH show that for the glass components: - the release rate evolution follows an Arrhenius law, - in acid PH, there is a selective dissolution, - in basic PH, there is a stoechiometric dissolution [fr

Silicate glasses

International Nuclear Information System (INIS)

Lutze, W.

1988-01-01

Vitrification of liquid high-level radioactive wastes has received the greatest attention, world-wide, compared to any other HLW solidification process. The waste form is a borosilicate-based glass. The production of phosphate-based glass has been abandoned in the western world. Only in the Soviet Union are phosphate-based glasses still being developed. Vitrification techniques, equipment and processes and their remote operation have been developed and studied for almost thirty years and have reached a high degree of technical maturity. Industrial demonstration of the vitrification process has been in progress since 1978. This chapter is a survey of world-wide research and development efforts in nuclear waste glasses and its production technology. The principal glasses considered are silicate glasses which contain boron, i.e., borosilicate glasses

Radiation effects in moist-air systems and the influence of radiolytic product formation on nuclear waste glass corrosion

International Nuclear Information System (INIS)

Wronkiewicz, D.J.; Bates, J.K.; Buck, E.C.; Hoh, J.C.; Emery, J.W.; Wang, L.M.

1997-07-01

Ionizing radiation may affect the performance of glass in an unsaturated repository site by interacting with air, water vapor, or liquid water to produce a variety of radiolytic products. Tests were conducted to examine the effects of radiolysis under high gas/liquid ratios. Results indicate that nitrate is the predominant radiolytic product produced following both gamma and alpha radiation exposure, with lesser amounts of nitrite and carboxylic acids. The formation of nitrogen acids during exposure to long-lived, alpha-particle-emitting transuranic elements indicates that these acids may play a role in influencing nuclear waste form reactions in a long-term unsaturated disposal scenario. Experiments were also conducted with samples that simulate the composition of Savannah River Plant nuclear waste glasses. Radiolytic product formation in batch tests (340 m -1 , 90 C) resulted in a small increase in the release rates of many glass components, such as alkali and alkaline earth elements, although silicon and uranium release rates were slightly reduced indicating an overall beneficial effect of radiation on waste form stability. The radiolytic acids increased the rate of ion exchange between the glass and the thin film of condensate, resulting in accelerated corrosion rates for the glass. The paragenetic sequence of alteration phases formed on both the irradiated and nonirradiated glass samples reacted in the vapor hydration tests matches closely with those developed during volcanic glass alteration in naturally occurring saline-alkaline lake systems. This correspondence suggests that the high temperatures used in these tests have not changed the underlying glass reaction mechanism relate to that which controls glass reactions under ambient surficial conditions

Radiation effects in moist-air systems and the influence of radiolytic product formation on nuclear waste glass corrosion

Energy Technology Data Exchange (ETDEWEB)

Wronkiewicz, D.J.; Bates, J.K.; Buck, E.C.; Hoh, J.C.; Emery, J.W. [Argonne National Lab., IL (United States). Chemical Technology Div.; Wang, L.M. [Univ. of New Mexico, Albuquerque, NM (United States). Dept. of Geology

1997-07-01

Ionizing radiation may affect the performance of glass in an unsaturated repository site by interacting with air, water vapor, or liquid water to produce a variety of radiolytic products. Tests were conducted to examine the effects of radiolysis under high gas/liquid ratios. Results indicate that nitrate is the predominant radiolytic product produced following both gamma and alpha radiation exposure, with lesser amounts of nitrite and carboxylic acids. The formation of nitrogen acids during exposure to long-lived, alpha-particle-emitting transuranic elements indicates that these acids may play a role in influencing nuclear waste form reactions in a long-term unsaturated disposal scenario. Experiments were also conducted with samples that simulate the composition of Savannah River Plant nuclear waste glasses. Radiolytic product formation in batch tests (340 m{sup {minus}1}, 90 C) resulted in a small increase in the release rates of many glass components, such as alkali and alkaline earth elements, although silicon and uranium release rates were slightly reduced indicating an overall beneficial effect of radiation on waste form stability. The radiolytic acids increased the rate of ion exchange between the glass and the thin film of condensate, resulting in accelerated corrosion rates for the glass. The paragenetic sequence of alteration phases formed on both the irradiated and nonirradiated glass samples reacted in the vapor hydration tests matches closely with those developed during volcanic glass alteration in naturally occurring saline-alkaline lake systems. This correspondence suggests that the high temperatures used in these tests have not changed the underlying glass reaction mechanism relate to that which controls glass reactions under ambient surficial conditions.

Effects of container material on PCT leach test results for high-level nuclear waste glasses

International Nuclear Information System (INIS)

Xing, S.B.; Pegg, I.L.

1994-01-01

A glass-based waste form used for the immobilization of high-level nuclear wastes should exhibit good resistance to aqueous corrosion since typically this is the primary process by which radionucleides could be released into the environment upon failure of other barriers. In the USA, the Waste Acceptance Product Specifications (WAPS) provides a set of requirements to ensure the consistency of the waste forms produced and specifies the Product Consistency Test (PCT) as a measure of relative chemical durability. While the PCT procedure permits usage of both Teflon and stainless steel vessels for testing of simulated development glasses, Teflon is not permitted for testing of production glasses due to radiative degradation. The results presented in this paper indicate that there are very significant differences between tests conducted in the two types of vessels due to the well-known permeability of Teflon to atmospheric carbon dioxide which results in lowering of the solution pH and a consequent reduction in the leach rate of silicate glasses. A wide range of nuclear waste glass compositions was subjected to the PCT procedure using both Teflon and stainless steel vessels. The magnitude of the effect (up to a factor of four for B, Na, Li concentrations) depends strongly on glass composition, therefore the isolated checks performed previously were inconclusive. The permeability to CO, of two types of Teflon vessels specified in the PCT procedure was directly measured using buffer solutions: ingress of CO, is linear in time, strongly pH-dependent, and was as high as 100 ppm after 7 days. In actual PCT tests in Teflon vessels, the total CO, content was 560 ppm after 87 days and 1930 ppm after one year

Assessment of lead tellurite glass for immobilizing electrochemical salt wastes from used nuclear fuel reprocessing

Energy Technology Data Exchange (ETDEWEB)

Riley, Brian J.; Kroll, Jared O.; Peterson, Jacob A.; Pierce, David A.; Ebert, William L.; Williams, Benjamin D.; Snyder, Michelle M. V.; Frank, Steven M.; George, Jaime L.; Kruska, Karen

2017-11-01

This paper provides an overview of research evaluating the use of lead tellurite glass as a waste form for salt wastes from electrochemical reprocessing of used nuclear fuel. The efficacy of using lead tellurite glass to immobilize three different salt compositions was evaluated: a LiCl-Li2O oxide reduction salt containing fission products from oxide fuel, a LiCl-KCl eutectic salt containing fission products from metallic fuel, and SrCl2. Physical and chemical properties of glasses made with these salts were characterized with X-ray diffraction, bulk density measurements, differential thermal analysis, chemical durability tests, scanning and transmission electron microscopies, and energy-dispersive X-ray spectroscopy. These glasses were found to accommodate high salt concentrations and have high densities, but further development is needed to improve chemical durability. (C) 2017 Published by Elsevier B.V.

Nuclear traces in glass

International Nuclear Information System (INIS)

Segovia A, M. de N.

1978-01-01

The charged particles produce, in dielectric materials, physical and chemical effects which make evident the damaged zone along the trajectory of the particle. This damaged zone is known as the latent trace. The latent traces can be enlarged by an etching of the detector material. This treatment attacks preferently the zones of the material where the charged particles have penetrated, producing concavities which can be observed through a low magnification optical microscope. These concavities are known as developed traces. In this work we describe the glass characteristics as a detector of the fission fragments traces. In the first chapter we present a summary of the existing basic theories to explain the formation of traces in solids. In the second chapter we describe the etching method used for the traces development. In the following chapters we determine some chatacteristics of the traces formed on the glass, such as: the development optimum time; the diameter variation of the traces and their density according to the temperature variation of the detector; the glass response to a radiation more penetrating than that of the fission fragments; the distribution of the developed traces and the existing relation between this ditribution and the fission fragments of 252 Cf energies. The method which has been used is simple and cheap and can be utilized in laboratories whose resources are limited. The commercial glass which has been employed allows the registration of the fission fragments and subsequently the realization of experiments which involve the counting of the traces as well as the identification of particles. (author)

Effect of different glasses in glass bonded zeolite

International Nuclear Information System (INIS)

Lewis, M.A.; Ackerman, J.P.; Verma, S.

1995-01-01

A mineral waste form has been developed for chloride waste salt generated during the pyrochemical treatment of spent nuclear fuel. The waste form consists of salt-occluded zeolite powders bound within a glass matrix. The zeolite contains the salt and immobilizes the fission products. The zeolite powders are hot pressed to form a mechanically stable, durable glass bonded zeolite. Further development of glass bonded zeolite as a waste form requires an understanding of the interaction between the glass and the zeolite. Properties of the glass that enhance binding and durability of the glass bonded zeolite need to be identified. Three types of glass, boroaluminosilicate, soda-lime silicate, and high silica glasses, have a range of properties and are now being investigated. Each glass was hot pressed by itself and with an equal amount of zeolite. MCC-1 leach tests were run on both. Soda-lime silicate and high silica glasses did not give a durable glass bonded zeolite. Boroaluminosilicate glasses rich in alkaline earths did bind the zeolite and gave a durable glass bonded zeolite. Scanning electron micrographs suggest that the boroaluminosilicate glasses wetted the zeolite powders better than the other glasses. Development of the glass bonded zeolite as a waste form for chloride waste salt is continuing

Modifier cation effects on (29)Si nuclear shielding anisotropies in silicate glasses.

Science.gov (United States)

Baltisberger, Jay H; Florian, Pierre; Keeler, Eric G; Phyo, Pyae A; Sanders, Kevin J; Grandinetti, Philip J

2016-07-01

We have examined variations in the (29)Si nuclear shielding tensor parameters of SiO4 tetrahedra in a series of seven alkali and alkaline earth silicate glass compositions, Cs2O·4.81 SiO2, Rb2O·3.96 SiO2, Rb2O·2.25 SiO2, K2O·4.48 SiO2, Na2O·4.74 SiO2, BaO·2.64 SiO2, and SrO·2.36 SiO2, using natural abundance (29)Si two-dimensional magic-angle flipping (MAF) experiments. Our analyses of these 2D spectra reveal a linear dependence of the (29)Si nuclear shielding anisotropy of Q((3)) sites on the Si-non-bridging oxygen bond length, which in turn depends on the cation potential and coordination of modifier cations to the non-bridging oxygen. We also demonstrate how a combination of Cu(2+) as a paramagnetic dopant combined with echo train acquisition can reduce the total experiment time of (29)Si 2D NMR measurements by two orders of magnitude, enabling higher throughput 2D NMR studies of glass structure. Copyright © 2016 Elsevier Inc. All rights reserved.

Modifier cation effects on 29Si nuclear shielding anisotropies in silicate glasses

Science.gov (United States)

Baltisberger, Jay H.; Florian, Pierre; Keeler, Eric G.; Phyo, Pyae A.; Sanders, Kevin J.; Grandinetti, Philip J.

2016-07-01

We have examined variations in the 29Si nuclear shielding tensor parameters of SiO4 tetrahedra in a series of seven alkali and alkaline earth silicate glass compositions, Cs2O · 4.81 SiO2, Rb2O · 3.96 SiO2, Rb2O · 2.25 SiO2, K2O · 4.48 SiO2, Na2O · 4.74 SiO2, BaO · 2.64 SiO2, and SrO · 2.36 SiO2, using natural abundance 29Si two-dimensional magic-angle flipping (MAF) experiments. Our analyses of these 2D spectra reveal a linear dependence of the 29Si nuclear shielding anisotropy of Q(3) sites on the Si-non-bridging oxygen bond length, which in turn depends on the cation potential and coordination of modifier cations to the non-bridging oxygen. We also demonstrate how a combination of Cu2+ as a paramagnetic dopant combined with echo train acquisition can reduce the total experiment time of 29Si 2D NMR measurements by two orders of magnitude, enabling higher throughput 2D NMR studies of glass structure.

Modelling the local atomic structure of molybdenum in nuclear waste glasses with ab initio molecular dynamics simulations.

Science.gov (United States)

Konstantinou, Konstantinos; Sushko, Peter V; Duffy, Dorothy M

2016-09-21

The nature of chemical bonding of molybdenum in high level nuclear waste glasses has been elucidated by ab initio molecular dynamics simulations. Two compositions, (SiO 2 ) 57.5 -(B 2 O 3 ) 10 -(Na 2 O) 15 -(CaO) 15 -(MoO 3 ) 2.5 and (SiO 2 ) 57.3 -(B 2 O 3 ) 20 -(Na 2 O) 6.8 -(Li 2 O) 13.4 -(MoO 3 ) 2.5 , were considered in order to investigate the effect of ionic and covalent components on the glass structure and the formation of the crystallisation precursors (Na 2 MoO 4 and CaMoO 4 ). The coordination environments of Mo cations and the corresponding bond lengths calculated from our model are in excellent agreement with experimental observations. The analysis of the first coordination shell reveals two different types of molybdenum host matrix bonds in the lithium sodium borosilicate glass. Based on the structural data and the bond valence model, we demonstrate that the Mo cation can be found in a redox state and the molybdate tetrahedron can be connected with the borosilicate network in a way that inhibits the formation of crystalline molybdates. These results significantly extend our understanding of bonding in Mo-containing nuclear waste glasses and demonstrate that tailoring the glass composition to specific heavy metal constituents can facilitate incorporation of heavy metals at high concentrations.

Current Understanding and Remaining Challenges in Modeling Long-Term Degradation of Borosilicate Nuclear Waste Glasses

International Nuclear Information System (INIS)

Vienna, John D.; Ryan, Joseph V.; Gin, Stephane; Inagaki, Yaohiro

2013-01-01

Chemical durability is not a single material property that can be uniquely measured. Instead it is the response to a host of coupled material and environmental processes whose rates are estimated by a combination of theory, experiment, and modeling. High-level nuclear waste (HLW) glass is perhaps the most studied of any material yet there remain significant technical gaps regarding their chemical durability. The phenomena affecting the long-term performance of HLW glasses in their disposal environment include surface reactions, transport properties to and from the reacting glass surface, and ion exchange between the solid glass and the surrounding solution and alteration products. The rates of these processes are strongly influenced and are coupled through the solution chemistry, which is in turn influenced by the reacting glass and also by reaction with the near-field materials and precipitation of alteration products. Therefore, those processes must be understood sufficiently well to estimate or bound the performance of HLW glass in its disposal environment over geologic time-scales. This article summarizes the current state of understanding of surface reactions, transport properties, and ion exchange along with the near-field materials and alteration products influences on solution chemistry and glass reaction rates. Also summarized are the remaining technical gaps along with recommended approaches to fill those technical gaps

Characteristics of colloids generated during the corrosion of nuclear waste glasses in groundwater

International Nuclear Information System (INIS)

Feng, X.; Buck, E.C.; Mertz, C.; Bates, J.K.; Cunnane, J.C.; Chaiko, D.

1993-10-01

Aqueous colloidal suspensions were generated by reacting nuclear waste glasses with groundwater at 90 degrees C at different ratios of the glass surface area to solution volume (S/V). The colloids have been characterized in terms of size, charge, identity, and stability with respect to salt concentration, pH, and time, by examination using dynamic light scattering, electrophoretic mobility, and transmission electron microscopy. The colloids are predominately produced by precipitation from solution, possibly with contribution from reacted layers that have spallated from the glass. These colloids are silicon-rich minerals. The colloidal suspensions agglomerate when the salinity of the solutions increase. The following implications for modeling the colloidal transport of contaminants have been derived from this study: (1) The sources of the colloids are not only solubility-limited real colloids and the pseudo colloids formed by adsorption of radionuclides onto a groundwater colloid, but also from the spalled surface layers of reacted waste glasses. (2) In a repository, the local environment is likely to be glass-reaction dominated and the salt concentration is likely to be high, leading to rapid colloid agglomeration and settling; thus, colloid transport may be insignificant. (3) If large volumes of groundwater contact the glass reaction site, the precipitated colloids may become resuspended, and colloid transport may become important. (4) Under most conditions, the colloids are negatively charged and will deposit readily on positively charged surfaces. Negatively charged surfaces will, in general, facilitate colloid stability and transport

Radiation effects in silicate glasses

International Nuclear Information System (INIS)

Bibler, N.E.; Howitt, D.G.

1988-01-01

The study of radiation effects in complex silicate glasses has received renewed attention because of their use in special applications such as high level nuclear waste immobilization and fiber optics. Radiation changes the properties of these glasses by altering their electronic and atomic configurations. These alterations or defects may cause dilatations or microscopic phase changes along with absorption centers that limit the optical application of the glasses. Atomic displacements induced in the already disordered structure of the glasses may affect their use where heavy irradiating particles such as alpha particles, alpha recoils, fission fragments, or accelerated ions are present. Large changes (up to 1%) in density may result. In some cases the radiation damage may be severe enough to affect the durability of the glass in aqueous solutions. In the paper, the authors review the literature concerning radiation effects on density, durability, stored energy, microstructure and optical properties of silicate glasses. Both simple glasses and complex glasses used for immobilization of nuclear waste are considered

Standard test method for determining liquidus temperature of immobilized waste glasses and simulated waste glasses

CERN Document Server

American Society for Testing and Materials. Philadelphia

2011-01-01

1.1 These practices cover procedures for determining the liquidus temperature (TL) of nuclear waste, mixed nuclear waste, simulated nuclear waste, or hazardous waste glass in the temperature range from 600°C to 1600°C. This method differs from Practice C829 in that it employs additional methods to determine TL. TL is useful in waste glass plant operation, glass formulation, and melter design to determine the minimum temperature that must be maintained in a waste glass melt to make sure that crystallization does not occur or is below a particular constraint, for example, 1 volume % crystallinity or T1%. As of now, many institutions studying waste and simulated waste vitrification are not in agreement regarding this constraint (1). 1.2 Three methods are included, differing in (1) the type of equipment available to the analyst (that is, type of furnace and characterization equipment), (2) the quantity of glass available to the analyst, (3) the precision and accuracy desired for the measurement, and (4) candi...

Rhenium solubility in borosilicate nuclear waste glass: implications for the processing and immobilization of technetium-99.

Science.gov (United States)

McCloy, John S; Riley, Brian J; Goel, Ashutosh; Liezers, Martin; Schweiger, Michael J; Rodriguez, Carmen P; Hrma, Pavel; Kim, Dong-Sang; Lukens, Wayne W; Kruger, Albert A

2012-11-20

The immobilization of technetium-99 ((99)Tc) in a suitable host matrix has proven to be a challenging task for researchers in the nuclear waste community around the world. In this context, the present work reports on the solubility and retention of rhenium, a nonradioactive surrogate for (99)Tc, in a sodium borosilicate glass. Glasses containing target Re concentrations from 0 to 10,000 ppm [by mass, added as KReO(4) (Re(7+))] were synthesized in vacuum-sealed quartz ampules to minimize the loss of Re from volatilization during melting at 1000 °C. The rhenium was found as Re(7+) in all of the glasses as observed by X-ray absorption near-edge structure. The solubility of Re in borosilicate glasses was determined to be ~3000 ppm (by mass) using inductively coupled plasma optical emission spectroscopy. At higher rhenium concentrations, additional rhenium was retained in the glasses as crystalline inclusions of alkali perrhenates detected with X-ray diffraction. Since (99)Tc concentrations in a glass waste form are predicted to be wastes, assuming Tc as Tc(7+) and similarities between Re(7+) and Tc(7+) behavior in this glass system.

X-ray tomography of feed-to-glass transition of simulated borosilicate waste glasses

Czech Academy of Sciences Publication Activity Database

Harris, W.H.; Guillen, D.P.; Kloužek, Jaroslav; Pokorný, P.; Yano, T.; Lee, S.; Schweiger, M. J.; Hrma, P.

2017-01-01

Roč. 100, č. 9 (2017), s. 3883-3894 ISSN 0002-7820 Institutional support: RVO:67985891 Keywords : borosilicate glass * computed tomography * glass melting * morphology * nuclear waste * X-ray Subject RIV: JH - Ceramics, Fire-Resistant Materials and Glass OBOR OECD: Ceramics Impact factor: 2.841, year: 2016

Corrosion mechanisms and behaviour of actinides in the 'R7T7' nuclear glass

International Nuclear Information System (INIS)

Fillet, Sylvie

1987-01-01

This research thesis reports the study of aqueous corrosion of the R7T7 nuclear glass and of the identified corrosion mechanisms in conditions of static lixiviation which are close to that expected during long term storage in a geological environment. More specifically, this work aims at assessing the durability of this glass which has been selected for the vitrification of solutions from pressurized water reactors. The main glass alteration phenomena have been studied. The first part addresses the study of the alteration of the glassy matrix, and aims at identifying corrosion mechanisms in various lixiviation conditions (high temperature, saturation). The second part addresses the action of different materials present in the environment on the glassy matrix by simulating as well as possible a storage case. Based on the obtained results, a mathematical model is developed to predict the glass behaviour on the long term. Finally, the glass confinement power with respect to actinides is studied [fr

Alpha self irradiation effects in nuclear borosilicate glass

International Nuclear Information System (INIS)

Peuget, S.; Roudil, D.; Deschanels, X.; Jegou, C.; Broudic, V.; Bart, J.M.

2004-01-01

The properties of actinide glasses are studied in the context of high-level waste management programs. Reprocessing high burnup fuels in particular will increase the minor actinide content in the glass package, resulting in higher cumulative alpha decay doses in the glass, and raising the question of the glass matrix behavior and especially its containment properties. The effect of alpha self-irradiation on the glass behavior is evaluated by doping the glass with a short-lived actinide ( 244 Cm) to reach in several years the alpha dose received by the future glass packages over several thousand years. 'R7T7' borosilicate glasses were doped with 3 different curium contents (0.04, 0.4 and 1.2 wt% 244 CmO 2 ). The density and mechanical properties of the curium-doped glasses were characterized up to 2. 10 18 α/g, revealing only a slight evolution of the macroscopic behavior of R7T7 glass in this range. The leaching behavior of curium-doped glass was also studied by Soxhlet tests. The results do not show any significant evolution of the initial alteration rate with the alpha dose. (authors)

Liquid-Liquid Phase Separation in Model Nuclear Waste Glasses: A Solid-State Double-Resonance NMR Study

Energy Technology Data Exchange (ETDEWEB)

Martineau, Ch.; Michaelis, V.K.; Kroeker, S. [Univ Manitoba, Dept Chem, Winnipeg, MB R3T 2N2 (Canada); Schuller, S. [CEA Valrho Marcoule, LDMC, SECM, DTCD, DEN, F-30207 Bagnols Sur Ceze (France)

2010-07-01

Double-resonance nuclear magnetic resonance (NMR) techniques are used in addition to single-resonance NMR experiments to probe the degree of mixing between network-forming cations Si and B, along with the modifier cations Cs{sup +} and Na{sup +} in two molybdenum-bearing model nuclear waste glasses. The double-resonance experiments involving {sup 29}Si in natural abundance are made possible by the implementation of a CPMG pulse-train during the acquisition period of the usual REDOR experiments. For the glass with lower Mo content, the NMR results show a high degree of Si-B mixing, as well as an homogeneous distribution of the cations within the borosilicate network, characteristic of a non-phase-separated glass. For the higher-Mo glass, a decrease of B-Si(Q{sup 4}) mixing is observed, indicating phase separation. {sup 23}Na and {sup 133}Cs NMR results show that although the Cs{sup +} cations, which do not seem to be influenced by the molybdenum content, are spread within the borate network, there is a clustering of the Na{sup +} cations, very likely around the molybdate units. The segregation of a Mo-rich region with Na{sup +} cations appears to shift the bulk borosilicate glass composition toward the metastable liquid liquid immiscibility region and induce additional phase separation. Although no crystallization is observed in the present case, this liquid liquid phase separation is likely to be the first stage of crystallization that can occur at higher Mo loadings or be driven by heat treatment. From this study emerges a consistent picture of the nature and extent of such phase separation phenomena in Mo-bearing glasses, and demonstrates the potential of double-resonance NMR methods for the investigation of phase separation in amorphous materials. (authors)

Leaching behavior of glass ceramic nuclear waste forms

International Nuclear Information System (INIS)

Lokken, R.O.

1981-11-01

Glass ceramic waste forms have been investigated as alternatives to borosilicate glasses for the immobilization of high-level radioactive waste at Pacific Northwest Laboratory (PNL). Three glass ceramic systems were investigated, including basalt, celsian, and fresnoite, each containing 20 wt % simulated high-level waste calcine. Static leach tests were performed on seven glass ceramic materials and one parent glass (before recrystallization). Samples were leached at 90 0 C for 3 to 28 days in deionized water and silicate water. The results, expressed in normalized elemental mass loss, (g/m 2 ), show comparable releases from celsian and fresnoite glass ceramics. Basalt glass ceramics demonstrated the lowest normalized elemental losses with a nominal release less than 2 g/m 2 when leached in polypropylene containers. The releases from basalt glass ceramics when leached in silicate water were nearly identical with those in deionized water. The overall leachability of celsian and fresnoite glass ceramics was improved when silicate water was used as the leachant

Proceedings of the national conference on functional glasses/glass-ceramics and ceramics: souvenir

International Nuclear Information System (INIS)

2015-01-01

This conference deals with issues relevant to functional glasses and glass ceramics which are technologically important materials for lasers, radioactive waste immobilization, radiation shielding, bio-glasses etc. It covers wide range of subjects and their applications right from managing the side effects of nuclear wastes and shielding the radiation, to sol-gel based bio-glass and its composites. Papers relevant to INIS are indexed separately

Thermal stability of the French nuclear waste glass - long term behavior modeling

International Nuclear Information System (INIS)

Orlhac, X.

2000-01-01

The thermal stability of the French nuclear waste glass was investigated experimentally and by modeling to predict its long-term evolution at low temperature. The crystallization mechanisms were analyzed by studying devitrification in the supercooled liquid. Three main crystalline phases were characterized (CaMoO 4 , CeCO 2 , ZnCr 2 O 4 ). Their crystallisation was TO 4.24 wt%, due to the low concentration of the constituent elements. The nucleation and growth curves showed that platinoid elements catalysed nucleation but did not affect growth, which was governed by volume diffusion. The criteria of classic nucleation theory were applied to determine the thermodynamic and diffusional activation energies. Viscosity measurements illustrate the analogy between the activation energy of viscous flow and diffusion, indicating control of crystallization by viscous flow phenomena. The combined action of nucleation and growth was assessed by TTT plots, revealing a crystallization equilibrium line that enables the crystallized fractions to be predicted over the long term. The authors show that hetero-genetics catalyze the transformation without modifying the maximum crystallized fraction. A kinetic model was developed to describe devitrification in the glass based on the nucleation and growth curves alone. The authors show that the low-temperature growth exhibits scale behavior (between time and temperature) similar to thermo-rheological simplicity. The analogy between the resulting activation energy and that of the viscosity was used to model growth on the basis of viscosity. After validation with a simplified (BaO 2 SiO 2 ) glass, the model was applied to the containment glass. The result indicated that the glass remained completely vitreous after a cooling scenario with the one measured at the glass core. Under isothermal conditions, several million years would be required to reach the maximum theoretical crystallization fraction. (author)

Crystallization In Multicomponent Glasses

International Nuclear Information System (INIS)

Kruger, A.A.; Hrma, P.R.

2009-01-01

In glass processing situations involving glass crystallization, various crystalline forms nucleate, grow, and dissolve, typically in a nonuniform temperature field of molten glass subjected to convection. Nuclear waste glasses are remarkable examples of multicomponent vitrified mixtures involving partial crystallization. In the glass melter, crystals form and dissolve during batch-to-glass conversion, melter processing, and product cooling. Crystals often agglomerate and sink, and they may settle at the melter bottom. Within the body of cooling glass, multiple phases crystallize in a non-uniform time-dependent temperature field. Self-organizing periodic distribution (the Liesegnang effect) is common. Various crystallization phenomena that occur in glass making are reviewed.

CRYSTALLIZATION IN MULTICOMPONENT GLASSES

Energy Technology Data Exchange (ETDEWEB)

KRUGER AA; HRMA PR

2009-10-08

In glass processing situations involving glass crystallization, various crystalline forms nucleate, grow, and dissolve, typically in a nonuniform temperature field of molten glass subjected to convection. Nuclear waste glasses are remarkable examples of multicomponent vitrified mixtures involving partial crystallization. In the glass melter, crystals form and dissolve during batch-to-glass conversion, melter processing, and product cooling. Crystals often agglomerate and sink, and they may settle at the melter bottom. Within the body of cooling glass, multiple phases crystallize in a non-uniform time-dependent temperature field. Self-organizing periodic distribution (the Liesegnang effect) is common. Various crystallization phenomena that occur in glass making are reviewed.

Aqueous corrosion of french R7T7 nuclear waste glass: selective then congruent dissolution by pH increase

International Nuclear Information System (INIS)

Advocat, T.; Vernaz, E.; Crovisier, J.L.

1991-01-01

A study of the corrosion of a borosilicate nuclear glass shows the strong effect of the pH on the dissolution mechanism. Acidic media lead to selective extraction of the glass modifier elements (Li, Na, Ca) as well as B, while dissolution is congruent under alkaline conditions. The silica dissolution rate significantly increases with increasing pH [fr

Influence of nuclear glasses composition on their liability to deterioration; Influence de la composition des verres nucleaires sur leur alterabilite

Energy Technology Data Exchange (ETDEWEB)

Tovena, I

1995-09-29

This contributes to the study of the nuclear glasses composition influence on their liability to deterioration. The methodology of the experimental research used has lead to define between the thirty oxides which form the reference glass light water, six oxides of interest. For each of these oxides, a composition variation area has been defined. A matrix of twenty glass compositions has then been defined. The preparation of materials of these compositions has sometimes lead to materials weakly heterogeneous which have been characterized before deterioration. This study has been completed by those of three glasses in a composition variation area narrower of the light water nuclear glass : the R7T7 and two glasses at limits having respectively an initial dissolution velocity at 100 degrees Celsius theoretically maximum and minimum. Some deterioration parameters in pure water have been experimentally measured on the twenty three glasses : 1) an initial dissolution velocity at 100 degrees (Vo{sub 1}00) Celsius and another one at 90 degrees Celsius (Vo{sub 9}0) 2) a dissolution velocity in conditions near the saturation at 90 degrees Celsius 3) an apparent solubility of glass based on the ortho silicic acid activity 4) the evolution of the dissolution kinetics at 90 degrees Celsius in sub-saturated medium towards saturated medium 5) the alteration films nature developed at the glasses surface during these last alteration tests. Some thermodynamic and structural models have been studied in order to predict Vo{sub 9}0 and Vo{sub 1}00. The dissolution kinetic law developed from reference glass dissolution results has been studied with the calculation code LIXIVER. It has not been able to be used for most of the glasses compositions studied. As a consequence, the glasses dissolution control by a surface reaction which are itself controlled by the only dissolved silica is an hypothesis which is not verified for the greater part of the glasses. (O.L.). refs., figs., tabs.

Heterogeneities in nuclear waste glass

International Nuclear Information System (INIS)

Ladirat, Ch.

1997-01-01

The industrial vitrification of high level radioactive wastes is a 2 stage process. During the first stage, the concentrated solution is heated in a spinning resistance oven at the temperature of 400 Celsius degrees till evaporation and calcination. The second stage begins when the dry residue falls into a melting pot that is maintained at a temperature of 1100-1150 Celsius degrees. Glass fretting is added and the glass is elaborated through the fusion of the different elements present in the melting pot. Heterogeneities in the glass may be associated to: - the presence in the solution to vitrify of insoluble elements from the dissolution of the fuel (RuO 2 , Rh, Pd), - the presence of minuscule metal scraps (Zr) that have been produced during the cutting of the fuel element, - the failures to conform to the technical specifications of the vitrification process, for instance, temperatures or flow rates when introducing the different elements in the melting pot. (A.C.)

The role of noble metals in electric melting of nuclear waste glass

International Nuclear Information System (INIS)

Roth, G.; Weisenburger, S.

1990-01-01

Electrical melting of nuclear waste glass in ceramic melters applies Joule heating, with the molten glass acting as the conductive medium. The local energy release inside the melt relieves from the restriction of external heat addition, allowing to scale up the melter to industrial units. Certainly, that principle makes the melter operation susceptible for changes of the electrical properties of the glass melt. Hence, the melt properties are required to be locally uniform and constant with time. Temporary fluctuations in the feed composition, however, are usually attenuated by the high retention times being in the order of a day and more. More essential for the melter operation are segregation effects occurring systematically. This behaviour can be observed in the case of the so-called noble metal elements Ruthenium, Palladium and Rhodium, belonging to the Platinum metal group. The subject of this paper is to describe the behaviour of the noble metals in electric melting and the problems they can contribute to. The discussion is based on detailed knowledge gained from PAMELA's LEWC processing and from large-scale vitrification of commercial-like waste simulate at INE/KfK. Finally, ways are indicated to solve the noble metal problem technically

Glass material oxidation and dissolution system: Converting miscellaneous fissile materials to glass

International Nuclear Information System (INIS)

Forsberg, C.W.; Ferrada, J.J.

1996-01-01

The cold war and the development of nuclear energy have resulted in significant inventories of miscellaneous fissile materials (MFMs). MFMs include (1) plutonium scrap and residue, (2) miscellaneous spent nuclear fuel (SNF), (3) certain hot cell wastes, and (4) many one-of-a-kind materials. Major concerns associated with the long-term management of these materials include: safeguards and nonproliferation issues; health, environment, and safety concerns. waste management requirements; and high storage costs. These issues can be addressed by converting the MFMs to glass for secure, long-term storage or repository disposal; however, conventional glass-making processes require oxide-like feed materials. Converting MFMs to oxide-like materials with subsequent vitrification is a complex and expensive process. A new vitrification process has been invented, the Glass Material Oxidation and Dissolution System (GMODS), which directly converts metals, ceramics, and amorphous solids to glass; oxidizes organics with the residue converted to glass; and converts chlorides to borosilicate glass and a secondary sodium chloride (NaCl) stream. Laboratory work has demonstrated the conversion of cerium (a plutonium surrogate), uranium, Zircaloy, stainless steel, multiple oxides, and other materials to glass. However, significant work is required to develop GMODS further for applications at an industrial scale. If implemented, GMODS will provide a new approach to manage these materials

Nepheline Crystallization in Nuclear Waste Glasses: Progress toward acceptance of high-alumina formulations

International Nuclear Information System (INIS)

McCloy, John S.; Schweiger, Michael J.; Rodriguez, Carmen P.; Vienna, John D.

2011-01-01

We have critically compiled and analyzed historical data for investigating the quantity of nepheline (NaAlSiO4) precipitation as a function of composition in simulated nuclear waste glasses. To understand composition we used two primary methods: (1) investigating the Al2O3-SiO2-Na2O ternary with filtering for different B2O3 levels and (2) creating a quadrant system consisting of compositions reduced to two metric numbers. These metrics are (1) the nepheline discriminator (ND) which depends only on the SiO2 content by weight normalized to the total weight of the Al2O3-SiO2-Na2O sub-mixture and (2) the optical basicity (OB) which contains contributions from all constituents in the glass. Nepheline precipitation is expected to be suppressed at high SiO2 levels (ND >0.62) or at low basicities (OB 5 wt% B2O3. The OB concept can help further refine regions of nepheline-free glass formation.

Development of basic data for modelling the residual alteration rate in aqueous media of AVM nuclear glasses

International Nuclear Information System (INIS)

Thien, B.

2010-01-01

During their aqueous alteration, AVM French nuclear glasses exhibit a large range of behaviour, in spite of a small range of composition. AVM glasses alteration rates are controlled by two phenomena: (i) precipitation of secondary phases, mostly aluminous hectorites, and (ii) diffusion of water across a more or less protective gel. The magnesium contained in these glasses increases the precipitation of these secondary phases, leading to a partial or total dissolution of the gel layer. This dissolution increases the glass alteration rates. On the other hand, Mg also incorporates in the gel, increasing his passivation properties. The predominance of one of these two phenomena depends on the initial composition of the glass, the pH of the solution, and the alteration conditions. In presence of Bure geological disposal site water (Mg and Ca rich), AVM glasses undergo less alteration than in initially pure water, in spite of larger amounts of secondary phase precipitates. This results from incorporation of calcium in the gel instead of sodium and magnesium, improving its passivating properties. We have adapted the geochemical GRAAL model for AVM glasses. In spite of its limitations, this model allows us to describe the differences of behaviour between these glasses, in function of their composition. Moreover, GRAAL can be proposed as a basis of a future operational model for predicting the alteration of AVM glasses. (author) [fr

Study of the mechanisms underlying resumptions of alteration. Modeling and evaluation of the impact on nuclear waste glasses

International Nuclear Information System (INIS)

Fournier, Maxime

2015-01-01

A sudden and still poorly understood phenomenon, the resumption of alteration results in a sudden acceleration of the glass alteration rate due to the destabilization of the amorphous passivating layer formed on the glass surface. Understanding the origin and the consequences of this phenomenon is a major issue for the prediction of nuclear glass long-term behavior. This study quantitatively links the alteration degree of a six-oxide reference glass and the formation mechanisms of zeolites and C-S-H that control the solution chemistry. The role played by the decrease in aluminum concentration as an indicator of resumption is highlighted. It appears that the resumption occurrence and rate are correlated to the couple (T, pH), but even in the most adverse situations the resumption rate is lower than the initial alteration rate, which remains the fastest kinetic regime. Previously limited to alkaline pH, the characterization of alteration resumptions was extended to conditions more representative of those found in a geological repository. This approach required the development of a new tool: seeding, that reduces or eliminates the latency period preceding a resumption. The results obtained demonstrate its usefulness in understanding the role of zeolites in amorphous layer destabilization and for modeling alteration resumptions. A geochemical modeling approach to alteration resumption is proposed, based on the formalism of the GRAAL glass alteration model. It is based on the calculation of zeolite thermodynamic constants, on the implementation of their nucleation and growth kinetics, and on assumptions related to the solubility of the amorphous layer. When zeolite precipitation consumes alkali, glass alteration - driven by zeolite precipitation - releases alkali. The model highlights the importance of such chemical couplings and shows that, in the stoichiometry of French nuclear reference glass, their cumulative effects are the cause of a pH decrease which limits

Origin and evolution of cup-shaped structures on leached nuclear waste containment glass surfaces

International Nuclear Information System (INIS)

Dubois, C.; Villa, F.; Chambaudet, A.; Vernaz, E.

1994-01-01

A three-dimensional surface microanalysis system equipped with a sensitive topographical probe was used to quantify the evolution of cup-shaped structures formed by aqueous leaching of nuclear waste containment glass. A model of the dissolution phenomenon provides satisfactory correlation between calculated and measured cup radius and depth. Dissolution cups form from cracks on the initially cut glass surface. Large cracks control the phenomenon by forming the largest cups, which gradually absorb smaller ones. The evolution of the size and shape of the dissolution cups was described by a model that assumes a constant dissolution rate on the surface, diminishing with crack depth. The best fit with the experimental data was obtained with a dissolution rate one hundred times lower at the bottom of the crack than at the surface. Moreover, it is predictable that all the cups will gradually disappear as they grow larger and flatter over a leaching period of some 2 years, for the glass composition and experimental leaching procedures used in this work

Radionuclides containment in nuclear glasses. An overview

Energy Technology Data Exchange (ETDEWEB)

Gin, Stephane; Jollivet, Patrick; Tribet, Magaly; Peuget, Sylvain; Schuller, Sophie [CEA Marcoule, Bagnols-sur-Ceze (France). DE2D SEVT

2017-07-01

Radioactive waste vitrification has been carried out industrially in several countries for nearly 40 years. Research into the formulation and long term behavior of high and intermediate level waste glasses, mainly borosilicate compositions, is still continuing in order to (i) safely condition new types of wastes and (ii) design and demonstrate the safety of the disposal of these long-lived waste forms in a deep geological repository. This article presents a summary of current knowledge on the formulation, irradiation resistance and the chemical durability of these conditioning materials, with a special focus on the fate of radionuclides during glass processing and aging. It is shown that, apart from the situation for certain elements with very low incorporation rate in glass matrices, vitrification in borosilicate glass can enable waste loadings of up to ∝20 wt% while maintaining the glass homogeneity for geological time scales and guaranteeing a high stability level in spite of irradiation and water contact.

Crystallization of nuclear glass under a thermal gradient: application to the self-crucible produced in the skull melting process

International Nuclear Information System (INIS)

Delattre, O.

2013-01-01

In the context of the vitrification of high level nuclear waste, a new industrial process has been launched in 2010 at the La Hague factory: The skull melting process. This setup applies thermal gradients to the melt, which leads to the formation of a solid layer of glass: the 'self-crucible'. The question would be to know whether these thermal gradients have an impact or not on the crystallization behaviour of the considered glasses in the self crucible. In order to answer that question, the crystallization of two glass compositions of nuclear interest has been investigated with an image analysis based method in isothermal and thermal gradient heat treatments conditions. The isothermal experiments allow for the quantification (growth speed, nucleation, crystallized fraction) of the crystallization of apatites (660 C-900 C) and powellites (630 C-900 C). The comparison of the results obtained through these two types of experimentations allows us to conclude that there is no impact of the thermal gradient on the crystallization of the studied glass compositions. In order to complete the image analysis study (based on surfaces), in and ex situ microtomography experiments have been performed at ESRF (Grenoble) on the ID19 beamline. This study allowed us to follow the crystallization of apatites in a simplified glass and to confirm the reliability of the image analysis method based on the analysis of surfaces. (author) [fr

Specific outcomes of the research on the radiation stability of the French nuclear glass towards alpha decay accumulation

Science.gov (United States)

Peuget, S.; Delaye, J.-M.; Jégou, C.

2014-01-01

This paper presents an overview of the main results of the French research on the long-term behavior of SON68 nuclear glass towards alpha decay accumulation. The effect of the radiation damage induced by alpha decay and also helium build-up were investigated by examining glass specimens, doped with a short-lived actinide 244Cm, irradiated by light and heavy ions. Additionally, atomistic simulations by molecular dynamics have provided further information on the atomic-scale effects of the macroscopic phenomena observed. These studies have shown that some macroscopic properties vary with the accumulation of alpha decay, but then stabilize after integrated doses of the order of 4 × 1018 α g-1. For example, the glass density diminishes by about 0.6%, its Young's modulus by about 15%, and its hardness by about 30%, while its fracture toughness increases by around 50%. The SEM and TEM characterization showed that the glass is still homogeneous. No phase separation, crystallization or bubbles formation was noticed up to an alpha decay dose corresponding to several thousand years of disposal of nuclear glass canister. Moreover the initial alteration rate of the glass is not significantly affected by the glass damage induced by alpha decays or heavy ions irradiations. The comparison of the macroscopic evolutions of the Cm doped glass with those obtained for glasses irradiated with light or heavy ions (from either experimental and molecular dynamic studies) suggests that the macroscopic evolutions are induced by the nuclear interactions induced by the recoil nuclei of alpha decay. The analysis of the behavior of the glass structure subjected to ballistic effects with various spectroscopic studies, together with the results of atomistic modeling by molecular dynamics, have identified some slight changes in the local order around some cations. Moreover a modification of the medium-range order has also been demonstrated through changes in the bond angles between network

Specific outcomes of the research on the radiation stability of the French nuclear glass towards alpha decay accumulation

Energy Technology Data Exchange (ETDEWEB)

Peuget, S., E-mail: sylvain.peuget@cea.fr; Delaye, J.-M.; Jégou, C.

2014-01-15

This paper presents an overview of the main results of the French research on the long-term behavior of SON68 nuclear glass towards alpha decay accumulation. The effect of the radiation damage induced by alpha decay and also helium build-up were investigated by examining glass specimens, doped with a short-lived actinide {sup 244}Cm, irradiated by light and heavy ions. Additionally, atomistic simulations by molecular dynamics have provided further information on the atomic-scale effects of the macroscopic phenomena observed. These studies have shown that some macroscopic properties vary with the accumulation of alpha decay, but then stabilize after integrated doses of the order of 4 × 10{sup 18} α g{sup −1}. For example, the glass density diminishes by about 0.6%, its Young’s modulus by about 15%, and its hardness by about 30%, while its fracture toughness increases by around 50%. The SEM and TEM characterization showed that the glass is still homogeneous. No phase separation, crystallization or bubbles formation was noticed up to an alpha decay dose corresponding to several thousand years of disposal of nuclear glass canister. Moreover the initial alteration rate of the glass is not significantly affected by the glass damage induced by alpha decays or heavy ions irradiations. The comparison of the macroscopic evolutions of the Cm doped glass with those obtained for glasses irradiated with light or heavy ions (from either experimental and molecular dynamic studies) suggests that the macroscopic evolutions are induced by the nuclear interactions induced by the recoil nuclei of alpha decay. The analysis of the behavior of the glass structure subjected to ballistic effects with various spectroscopic studies, together with the results of atomistic modeling by molecular dynamics, have identified some slight changes in the local order around some cations. Moreover a modification of the medium-range order has also been demonstrated through changes in the bond angles

Study of the surface crystallization and resistance to dissolution of niobium phosphate glasses for nuclear waste immobilization

International Nuclear Information System (INIS)

Vieira, Heveline

2008-01-01

The surface crystallization and the dissolution rate of three phosphate glass compositions containing different amounts of niobium oxide were studied. The glasses were named Nb30, Nb37, and Nb44 according to the nominal content of niobium oxide in the glass composition. The three compositions were evaluated keeping the P 2 O 5 /K 2 O ratio constant and varying the amount of Nb 2 O 5 . These glasses were produced by melting appropriate chemical compounds at 1500 deg C for 0.5 hour. The crystalline phases which were nucleated on the glass surface after heat treatment were determined by X-ray diffraction. The crystalline structures depend on the amount of niobium oxide in the glass composition. The crystal morphologies were observed by using an optical microscope, and their characteristics are specific for each kind of crystalline phase. The crystal growth rate and the surface nuclei density were determined for each glass composition, and they depend on each crystalline phase nucleated on the surface. From the differential thermal analysis curves it was determined that the Nb44 glass containing 46.5 mol por cent of niobium oxide is the most thermally stable against crystallization when compared to the Nb30 and Nb37 glasses. According to the activation energies determined for crystal growth on the surface of each glass type, the Nb44 glass can also be considered the most resistant one against crystallization. The dissolution rate for the Nb44 glass after 14 days immersed in an aqueous solution with pH equals to 7 at 90 deg C is the lowest (9.0 x 10 -7 g. cm -2 . day -1 ) when compared to the other two glass compositions. The dissolution rates in acidic and neutral solutions of all studied glasses meet the international standards for materials which can be used in the immobilization of nuclear wastes. (author)

Glass binder development for a glass-bonded sodalite ceramic waste form

International Nuclear Information System (INIS)

Riley, Brian J.; Vienna, John D.; Frank, Steven M.; Kroll, Jared O.; Peterson, Jacob A.

2017-01-01

This paper discusses work to develop Na_2O-B_2O_3-SiO_2 glass binders for immobilizing LiCl-KCl eutectic salt waste in a glass-bonded sodalite waste form following electrochemical reprocessing of used metallic nuclear fuel. In this paper, five new glasses with ~20 mass% Na_2O were designed to generate waste forms with high sodalite. The glasses were then used to produce ceramic waste forms with a surrogate salt waste. The waste forms made using these new glasses were formulated to generate more sodalite than those made with previous baseline glasses for this type of waste. The coefficients of thermal expansion for the glass phase in the glass-bonded sodalite waste forms made with the new binder glasses were closer to the sodalite phase in the critical temperature region near and below the glass transition temperature than previous binder glasses used. Finally, these improvements should result in lower probability of cracking in the full-scale monolithic ceramic waste form, leading to better long-term chemical durability.

Corrosion of synthesized glasses and glazes as analogs for nuclear waste glass degradation

International Nuclear Information System (INIS)

Vandiver, P.B.

1994-01-01

Synthesized glasses provide an opportunity to study natural corrosion processes which are intermediate in time span between geological examples of natural glasses, such as obsidians and tektites, and relatively short term laboratory tests lasting a few hours to several decades. In addition, synthesized glasses can usually be tracked to particular archaeological find sites with known dates of production and often burial. Environmental conditions are routinely measured at archaeological sites as a part of the excavation-process, such that information is available on the yearly cycling of temperature and relative humidity, sometimes at the depth at which the artifact was found. Whether the artifacts were excavated in an air enclosure, such as a tomb, or in the soil can also be reconstructed, such that one can determine whether aqueous or atmospheric corrosion was involved in the degradation process. For instance, so-called open-quotes Roman glassclose quotes may span a time period of production of 800 years and a geographical range from Germany to North Africa and from Britain to Afghanistan. One example is the storage during World War II of glass from the British Museum in underground metro stations. Some of these glasses have been in collections for over 100 years. Thus, populations of glasses can be chosen for experimentation which compare variations in bulk composition, dopants, microstructure, heat treatment, ground vs. fire polished surfaces, aqueous vs. atmospheric corrosion, geographic, geological as well as recent storage conditions. Glasses in museums are generally considered to have had their corrosion arrested and be stable because changes in visual appearance are not obvious. However, if we attempt to measure the range of surface water content in these glasses using Fourier transform infrared analysis, a considerable variability is found, as shown

Defect production in silica glasses under gamma-irradiation at the quenched nuclear reactor

International Nuclear Information System (INIS)

Mussaeva, M.A.; Kalanov, M.U.; Ibragimova, E.M.; Sandalov, V.N.; Muminov, M.L.

2004-01-01

Full text: Radiation defect production in oxides is highly interesting for atom and solar energy, and also for burying nuclear waste. Combine effect of neutron and gamma-radiation on materials was studied extensively and only neutrons are believed to displace atoms, although 60 Co-gamma quanta were proved to displace light anions (O, F) by inelastic mechanism. On the example of polished plates of pure fused quartz and barium-silica glasses containing nano-crystalline inclusions, and also nano-porous glass, the effect of gamma-radiation of the quenched reactor was studied in the energy range of 0.2-7 MeV. The time period was selected when practically constant current ∼10-20 nA is maintained in the ionizing chamber, corresponding to the average gamma-flux of 15-30 Gy/s. Optical absorption and photoluminescence spectra and also structure of the grasses were studied. It turned out, that the charged oxygen vacancies accumulation rate is higher in Barium glass than in the pure one, because for SiO 2 with small Z the photoelectric effect is weak, while the Compton scattering and photonuclear reactions prevail, and for Barium - just the opposite. The radiation-induced growth of the crystalline precipitates was noticed in the both glasses, which before had been attributed to the elastic atom displacements by fast neutrons. The density of Ba-glass increases with irradiation. The efficiency of defect production by the gamma-component even of the quenched reactor turned out much higher than that under irradiation with 60 Co gamma-source of ∼1.25 MeV to the equivalent dose at the current dose rate of ∼ 7 Gy/s (and before at 45 Gy/s). A 100-times increase of the surface proton conductivity was discovered in the porous glasses under gamma-irradiation due to water vapor radiolysis on the pore surface. The irradiated porous glass is recommended as an active electrode in the hydrogen fuel element. The work was done under the grant F2.1.2 from Center of Science and Technology

Glass and Glass-Ceramic Materials from Simulated Composition of Lunar and Martian Soils: Selected Properties and Potential Applications

Science.gov (United States)

Ray, C. S.; Sen, S.; Reis, S. T.; Kim, C. W.

2005-01-01

In-situ resource processing and utilization on planetary bodies is an important and integral part of NASA's space exploration program. Within this scope and context, our general effort is primarily aimed at developing glass and glass-ceramic type materials using lunar and martian soils, and exploring various applications of these materials for planetary surface operations. Our preliminary work to date have demonstrated that glasses can be successfully prepared from melts of the simulated composition of both lunar and martian soils, and the melts have a viscosity-temperature window appropriate for drawing continuous glass fibers. The glasses are shown to have the potential for immobilizing certain types of nuclear wastes without deteriorating their chemical durability and thermal stability. This has a direct impact on successfully and economically disposing nuclear waste generated from a nuclear power plant on a planetary surface. In addition, these materials display characteristics that can be manipulated using appropriate processing protocols to develop glassy or glass-ceramic magnets. Also discussed in this presentation are other potential applications along with a few selected thermal, chemical, and structural properties as evaluated up to this time for these materials.

Immobilization of high level nuclear wastes in sintered glasses. Devitrification evaluation produced with different thermal treatments

International Nuclear Information System (INIS)

Messi de Bernasconi, N.B.; Russo, D.O.; Bevilacqua, M.E.; Sterba, M.E.; Heredia, A.D.; Audero, M.A.

1990-01-01

This work describes immobilization of high level nuclear wastes in sintered glass, as alternative way to melting glass. Different chemical compositions of borosilicate glass with simulate waste were utilized and satisfactory results were obtained at laboratory scale. As another contribution to the materials studies by X ray powder diffraction analysis, the devitrification produced with different thermal treatments, was evaluated. The effect of the thermal history on the behaviour of fission products containing glasses has been studied by several working groups in the field of high level waste fixation. When the glass is cooled through the temperature range from 800 deg C down to less than 400 deg C (these temperatures are approximates) nucleation and crystal growth can take place. The rate of crystallization will be maximum near the transformation point but through this rate may be low at lower temperatures, devitrification can still occur over long periods of time, depending on the glass composition. It was verified that there can be an appreciable increase in leaching in some waste glass compositions owing to the presence of crystalline phases. On the other hand, other compositions show very little change in leachability and the devitrified product is often preferable as there is less tendency to cracking, particularly in massive blocks of glass. A borosilicate glass, named SG7, which was developed specially in the KfK for the hot pressing of HLW with glass frit was studied. It presents a much enhanced chemical durability than borosolicate glass developed for the melting process. The crystallization behaviour of SG7 glass products was investigated in our own experiments by annealing sintered samples up to 3000 h at temperatures between 675 and 825 deg C. The samples had contained simulated waste with noble metals, since these might act as foreign nuclei for crystallization. Results on the extent of devitrification and time- temperature- transformation curves are

Hydration of high-silica glasses in the deep sea

International Nuclear Information System (INIS)

Federman, A.N.

1986-01-01

Natural analogs of nuclear waste glasses are important because they provide information of the one variable that is not controllable in the laboratory - long intervals of time in the actual environment of storage. Some natural glasses have persisted for millions of years in deep-sea sediments in the form of disseminated particles and distinct tephra layers, while other apparently similar specimens have been completely altered to clay assemblages relatively quickly. Geologists have reached no firm conclusions as to why these differences exist, and more research is certainly warranted. These glasses vary in age, composition, and in the in-situ conditions they have experienced. They may provide important information for two different aspects of nuclear waste glass research: First, the chemical composition and especially the water content of these glasses as a function of time may give an understanding of the mechanisms and rates of diffusion in glasses in the natural environment. Second, the apparent differing durability of these glasses in different environmental conditions may suggest the optimal characteristics of a nuclear waste glass depository

Immobilization of Uranium Silicides in Sintered Glass

International Nuclear Information System (INIS)

Mateos, P.; Russo, D.O.; Heredia, A.D.; Sanfilippo, M.

2003-01-01

High activity nuclear spent fuels vitrification by fusion is a well known technology which has industrial scale in France, England, Japan, EEUU. Borosilicates glasses are used in this process.Sintered glasses are an alternative to the immobilization task in which there is also a wide experience around the world.The available technics are: cold pressing and sintering , hot-pressing and hot isostatic pressing.This work compares Borosilicates and Iron silicates sintered glasses behaviour when different ammounts of nuclear simulated waste is added

Reaction and devitrification of a prototype nuclear-waste-storage glass with hot magnesium-rich brine

International Nuclear Information System (INIS)

Komarneni, S.; Freeborn, W.P.; Scheetz, B.E.; White, W.B.; McCarthy, G.J.

1982-10-01

PNL 76-68, a prototype nuclear waste storage glass, was reacted under hydrothermal conditions at 100, 200, and 300 C with NBT-6a (Ca-Mg-K-Na-Cl) brine. Reaction products were identified, the state of the residual glass determined, and the concentrations of various elements remaining in the solutions analyzed. Solid products formed by reaction of the glass and brine talc (hydrated magnesium silicate), powellite (CaMoO 4 ), hematite (Fe 2 O 3 ) and rarely an unidentified uranium-containing phase. Glass fragments were leached to depths of 300 to 500 μm, depending on time and temperature. Most elements were extracted, but the silicate framework remained intact. Distinct diffusion fronts due to K/Na exchange and Mg/Zn exchange were identified. A complex compositional layering develops in the outer reaction rind. The concentration of silica in brine solution was lower by an order of magnitude than the concentration of silica in deionized water reacted under similar conditions. The concentration of cesium, strontium, uranium, rare earths, and other alkali and alkaline earth elements in solution increases exponentially with temperature of reaction. Behavior of the transition metals is more complex. In general the extraction of elements from the glass by hydrothermal brine leads to concentrations in solution that are from 10 to 100 times higher than the concentrations obtained by deionized water extraction under similar conditions of temperature and pressure

Effect of Callovo-Oxfordian clay rock on the dissolution rate of the SON68 simulated nuclear waste glass

Energy Technology Data Exchange (ETDEWEB)

Neeway, James J., E-mail: James.Neeway@pnnl.gov [SUBATECH, Unité Mixte de Recherche 6457, École des Mines de Nantes, CNRS/IN2P3, Université de Nantes, BP 20722, 44307 Nantes cedex 3 (France); Abdelouas, Abdesselam; Ribet, Solange; El Mendili, Yassine [SUBATECH, Unité Mixte de Recherche 6457, École des Mines de Nantes, CNRS/IN2P3, Université de Nantes, BP 20722, 44307 Nantes cedex 3 (France); Schumacher, Stéphan [ANDRA, Parc de la Croix Blanche, 1/7 rue Jean Monnet, 92298 Châtenay-Malabry (France); Grambow, Bernd [SUBATECH, Unité Mixte de Recherche 6457, École des Mines de Nantes, CNRS/IN2P3, Université de Nantes, BP 20722, 44307 Nantes cedex 3 (France)

2015-04-15

Long-term storage of high-level nuclear waste glass in France is expected to occur in an engineered barrier system (EBS) located in a subsurface Callovo-Oxfordian (COx) clay rock formation in the Paris Basin in northeastern France. Understanding the behavior of glass dissolution in the complex system is critical to be able to reliably model the performance of the glass in this complex environment. To simulate this multi-barrier repository scenario in the laboratory, several tests have been performed to measure glass dissolution rates of the simulated high-level nuclear waste glass, SON68, in the presence of COx claystone at 90 °C. Experiments utilized a High-Performance Liquid Chromatography (HPLC) pump to pass simulated Bure site COx pore water through a reaction cell containing SON68 placed between two COx claystone cores for durations up to 200 days. Silicon concentrations at the outlet were similar in all experiments, even the blank experiment with only the COx claystone (∼4 mg/L at 25 °C and ∼15 mg/L at 90 °C). The steady-state pH of the effluent, measured at room temperature, was roughly 7.1 for the blank and 7.3–7.6 for the glass-containing experiments demonstrating the pH buffering capacity of the COx claystone. Dissolution rates for SON68 in the presence of the claystone were elevated compared to those obtained from flow-through experiments conducted with SON68 without claystone in silica-saturated solutions at the same temperature and similar pH values. Additionally, through surface examination of the monoliths, the side of the monolith in direct contact with the claystone was seen to have a corrosion thickness 2.5× greater than the side in contact with the bulk glass powder. Results from one experiment containing {sup 32}Si-doped SON68 also suggest that the movement of Si through the claystone is controlled by a chemically coupled transport with a Si retention factor, K{sub d}, of 900 mL/g.

Retention of Halogens in Waste Glass

Energy Technology Data Exchange (ETDEWEB)

Hrma, Pavel R.

2010-05-01

In spite of their potential roles as melting rate accelerators and foam breakers, halogens are generally viewed as troublesome components for glass processing. Of five halogens, F, Cl, Br, I, and At, all but At may occur in nuclear waste. A nuclear waste feed may contain up to 10 g of F, 4 g of Cl, and ≤100 mg of Br and I per kg of glass. The main concern is halogen volatility, producing hazardous fumes and particulates, and the radioactive iodine 129 isotope of 1.7x10^7-year half life. Because F and Cl are soluble in oxide glasses and tend to precipitate on cooling, they can be retained in the waste glass in the form of dissolved constituents or as dispersed crystalline inclusions. This report compiles known halogen-retention data in both high-level waste (HLW) and low-activity waste (LAW) glasses. Because of its radioactivity, the main focus is on I. Available data on F and Cl were compiled for comparison. Though Br is present in nuclear wastes, it is usually ignored; no data on Br retention were found.

Composition effects on chemical durability and viscosity of nuclear waste glasses - systematic studies and structural thermodynamic models

International Nuclear Information System (INIS)

Feng, X.

1988-01-01

Two of the primary criteria for the acceptability of nuclear waste glasses are their durability, i.e. chemical resistance to aqueous attack for 10 4 to 10 5 years, and processability, which requires their viscosity at the desired melt temperature to be sufficiently low. Chapter 3 presents the results of systematic composition variation studies around the preliminary reference glass composition WV205 and an atomistic interpretation of the effects of individual oxides. Chapter 4 is concerned with modifications of the Jantzen-Plodinec hydration model which takes into account formation of complex aluminosilicate compounds in the glass. Chapter 5 is devoted to the development and validation of the structural-thermodynamic model for both durability and viscosity. This model assumes the strength of bonds between atoms to be the controlling factor in the composition dependence of these glass properties. The binding strengths are derived from the known heats of formation and the structural roles of constituent oxides. Since the coordination state of various oxides in the glass is temperature dependent and cation size has opposite effects on the two properties, the correlation between melt viscosity and rate of corrosion at low temperature is not simply linear. Chapter 6 surveys the effects of aqueous phase composition on the leach behavior of glasses. These studies provide a comprehensive view of the effects of both glass composition and leachant composition on leaching. The models developed correlate both durability and viscosity with glass composition. A major implication is that these findings can be used in the systematic optimization of the properties of complex oxide glasses

Basaltic glasses from Iceland and the deep sea: Natural analogues to borosilicate nuclear waste-form glass

International Nuclear Information System (INIS)

Jercinovic, M.J.; Ewing, R.C.

1987-12-01

The report provides a detailed analysis of the alteration process and products for natural basaltic glasses. Information of specific applicability to the JSS project include: * The identification of typical alteration products which should be expected during the long-term corrosion process of low-silica glasses. The leached layers contain a relatively high proportion of crystalline phases, mostly in the form of smectite-type clays. Channels through the layer provide immediate access of solutions to the fresh glass/alteration layer interface. Thus, glasses are not 'protected' from further corrosion by the surface layer. * Corrosion proceeds with two rates - an initial rate in silica-undersaturated environments and a long-term rate in silica-saturated environments. This demonstrates that there is no unexpected change in corrosion rate over long periods of time. The long-term corrosion rate is consistent with that of borosilicate glasses. * Precipitation of silica-containing phases can result in increased alteration of the glass as manifested by greater alteration layer thicknesses. This emphasizes the importance of being able to predict which phases form during the reaction sequence. * For natural basaltic glasses the flow rate of water and surface area of exposed glass are critical parameters in minimizing glass alteration over long periods of time. The long-term stability of basalt glasses is enhanced when silica concentrations in solution are increased. In summary, there is considerable agreement between corrosion phenomena observed for borosilicate glasses in the laboratory and those observed for natural basalt glasses of great age. (With 121 refs.) (authors)

The structure of phosphate glass biomaterials from neutron diffraction and 31P nuclear magnetic resonance data

International Nuclear Information System (INIS)

Pickup, D M; Ahmed, I; Guerry, P; Knowles, J C; Smith, M E; Newport, R J

2007-01-01

Neutron diffraction and 31 P nuclear magnetic resonance spectroscopy were used to probe the structure of phosphate glass biomaterials of general composition (CaO) 0.5-x (Na 2 O) x (P 2 O 5 ) 0.5 (x = 0, 0.1 and 0.5). The results suggest that all three glasses have structures based on chains of Q 2 phosphate groups. Clear structural differences are observed between the glasses containing Na 2 O and CaO. The P-O bonds to bridging and non-bridging oxygens are less well resolved in the neutron data from the samples containing CaO, suggesting a change in the nature of the bonding as the field strength of the cation increases Na + → Ca 2+ . In the (CaO) 0.5 (P 2 O 5 ) 0.5 glass most of the Ca 2+ ions are present in isolated CaO x polyhedra whereas in the (Na 2 O) 0.5 (P 2 O 5 ) 0.5 glass the NaO x polyhedra share edges leading to a Na-Na correlation. The results of the structural study are related to the properties of the (CaO) 0.4 (Na 2 O) 0.1 (P 2 O 5 ) 0.5 biomaterial

The dual effect of Mg on the long-term alteration rate of AVM nuclear waste glasses

Energy Technology Data Exchange (ETDEWEB)

Thien, Bruno M.J., E-mail: bruno.thien@psi.ch [Commissariat a l' Energie Atomique (CEA), Laboratoire d' Etude du Comportement a Long Terme des Materiaux, DTCD/DEN, Marcoule, 30207 Bagnols sur Ceze (France); Godon, Nicole; Ballestero, Anthony; Gin, Stephane [Commissariat a l' Energie Atomique (CEA), Laboratoire d' Etude du Comportement a Long Terme des Materiaux, DTCD/DEN, Marcoule, 30207 Bagnols sur Ceze (France); Ayral, Andre [Institut Europeen des Membranes, Universite de Montpellier, cc. 047, Place Eugene Bataillon, 34095 Montpellier (France)

2012-08-15

Inactive Mg-containing nuclear waste glasses simulating actual HLW glasses produced at the AVM facility since 1995 (Marcoule, France), were leached in aqueous solution in order to assess their long term behaviour. The focus was on the effect of Mg. Our findings show that the distribution of Mg between the gel and the secondary crystalline phases strongly influences the glass dissolution rate. The glasses were leached in initially pure water (T = 50 Degree-Sign C, surface/volume ratio (S/V) = 55 cm{sup -1}) with and without addition of Mg{sup 2+} in the solution. 'Mg-free' AVM glasses were also leached in initially pure water (50 Degree-Sign C, 200 cm{sup -1}) with and without addition of Mg{sup 2+} in the solution. Accurate identification of Mg-smectite secondary phases and gel composition calculations enable us to explain the different observed behaviours. Glass AVM 10 was the less altered glass in pure water. Its gel is more protective than the other probably because it is mainly balanced by Mg{sup 2+}. The addition of Mg{sup 2+} in the solution triggers the precipitation of smectite (not observed in pure water experiments), which consumes silicon from the gel, leading finally to a significant increase of the glass alteration. We also focused on the AVM 6 glass which was the most altered glass in pure water of available AVM glasses. Contrary to AVM 10, the gel of AVM 6 is mainly balanced by Na{sup +}. The addition of Mg{sup 2+} in the solution allows the replacement of Na by Mg within the gel. This reaction clearly improves the gel properties and allows the rate to decrease more rapidly, in spite of the precipitation of smectite (also observed in pure water experiments). Finally, the two glasses were altered in synthetic groundwater (SGW) with a high Mg-Ca content. As expected from the previous observations, AVM 10 was insensitive to the presence of alkaline earths in the leaching solution whereas AVM 6 glass exhibited a lower rate than in pure water

Waste glass weathering

International Nuclear Information System (INIS)

Bates, J.K.; Buck, E.C.

1994-01-01

The weathering of glass is reviewed by examining processes that affect the reaction of commercial, historical, natural, and nuclear waste glass under conditions of contact with humid air and slowly dripping water, which may lead to immersion in nearly static solution. Radionuclide release data from weathered glass under conditions that may exist in an unsaturated environment are presented and compared to release under standard leaching conditions. While the comparison between the release under weathering and leaching conditions is not exact, due to variability of reaction in humid air, evidence is presented of radionuclide release under a variety of conditions. These results suggest that both the amount and form of radionuclide release can be affected by the weathering of glass

Introduction to the crystallization phenomenon in nuclear glass

International Nuclear Information System (INIS)

Jacquet Francillon, N.

1997-01-01

Crystallization is a subject for concern because of its potentially detrimental effects on the technological feasibility of high-temperature melting, and on the chemical durability of the material at intermediate and low temperatures during interim storage or after disposal. The tendency of glass to crystallize depends to a large extent on the composition of the frit and/or of the waste to be solidified. It depends too of the thermal history of the glass generally, the knowledge is mainly upon determination of the time-temperature-transition (TTT) curves, crystal identification and quantification techniques, and their effects on the durability of the glass matrix. French experience is presented. Only a few authors addressed the long-term development of crystalline phases, notably at temperatures below the vitreous transition temperature Tg. Some recommendations for glass crystallization studies are made but glass crystallization after disposal is acceptable provided some conditions are met. (author)

Sulphate solubility and sulphate diffusion in oxide glasses: implications for the containment of sulphate-bearing nuclear wastes

International Nuclear Information System (INIS)

Lenoir, M.

2009-09-01

The thesis deals with sulphate solubility and sulphate diffusion in oxide glasses, in order to control sulphate incorporation and sulphate volatilization in nuclear waste glasses. It was conducted on simplified compositions, in the SiO 2 -B 2 O 3 -R 2 O (R = Li, Na, K, Cs), SiO 2 -B 2 O 3 -BaO and V 2 O 5 -B 2 O 3 -BaO systems. These compositions allowed us to study the influence of the nature of network-modifying ions (Li + , Na + , K + , Cs + or Ba 2+ ) and also of former elements (Si, B, V), on structure and properties of glasses. Sulphate volatility is studied in sodium borosilicate melts using an innovative technique of sulphate quantitation with Raman spectroscopy. This technique is useful to obtain kinetic curves of sulphate volatilization. The establishment of a model to fit these curves leads to the determination of diffusion coefficients of sulphate. These diffusion coefficients can thus be compared to diffusion coefficients of other species, determined by other techniques and presented in the literature. They are also linked to diffusion coefficients in relation with the viscosity of the melts. Concerning sulphate solubility in glasses, it depends on glass composition and on the nature of sulphate incorporated. Sulphate incorporation in alkali borosilicate glasses leads to the formation of a sulphate layer floating on top of the melt. Sulphate incorporation in barium borosilicate and boro-vanadate glasses leads to the crystallization of sulphate species inside the vitreous matrix. Moreover, sulphate solubility is higher in these glasses than in alkali borosilicates. Finally, exchanges between cations present in glasses and cations present in the sulphate phase are also studied. (author)

Effect of Ba in the glass characteristics of cesium loaded iron phosphate glasses

International Nuclear Information System (INIS)

Joseph, Kitheri; Asuvathraman, R.; Vasudeva Rao, P.R.

2015-01-01

Radioactive 137 Cs extracted from high level nuclear waste, when immobilized in a suitable matrix can be used as a γsource in medical industry. Iron phosphate glass (IPG) is one of a suitable matrix for the immobilization of 137 Cs prior to the immobilization of 137 Cs in IPG, it is essential to optimize the immobilization conditions using natural (inactive) cesium. Glass characteristics of inactive Cs loaded iron phosphate glasses were already explored in our earlier studies. However, the change in glass characteristics of 137 Cs loaded iron phosphate glass to 137 Ba loaded iron phosphate glass need to be studied before the immobilization of 137 Cs in iron phosphate glass as 137 Cs transforms to 137 Ba due to nuclear transmutation ( 137 Cs(β,γ) 137 Ba). This paper reports the studies on such a behaviour by incorporating inactive Ba in cesium loaded iron phosphate glasses. Cs and Ba loaded iron phosphate glasses were prepared by melt quench technique in air using appropriate amounts of Fe 2 O 3 , NH 4 H 2 PO 4 , Ba(OH) 2.8 H 2 O and Cs 2 CO 3 . The chemicals were added such that the glass formed possesses the batch composition of (a) 21.4 wt. % Fe 2 O 3 -45 wt. % Cs 2 O-5 wt % BaO-P 2 O 5 (henceforth referred as IP50Cs45Ba5); (b) 21.4 wt. % Fe 2 O 3 -25 wt. % Cs 2 O-25 wt % BaO-P 2 O5 (henceforth referred as IP50Cs25Ba25). The thermal expansion measurements were also carried out using a home-built quartz push-rod dilatometer. The data related to change in thermal expansion behaviour, glass forming ability, glass stability and structural changes in phosphate network due to the partial replacement of Cs with Ba will also be discussed. (author)

Heavy metal oxide glasses as gamma rays shielding material

International Nuclear Information System (INIS)

Kaur, Preet; Singh, Devinder; Singh, Tejbir

2016-01-01

The gamma rays shielding parameters for heavy metal oxide glasses and concrete samples are comparable. However, the transparent nature of glasses provides additional feature to visualize inside the shielding material. Hence, different researchers had contributed in computing/measuring different shielding parameters for different configurations of heavy metal oxide glass systems. In the present work, a detailed study on different heavy metal (_5_6Ba, _6_4Gd, _8_2Pb, _8_3Bi) oxide glasses has been presented on the basis of different gamma rays shielding parameters as reported by different researchers in the recent years. It has been observed that among the selected heavy metal oxide glass systems, Bismuth based glasses provide better gamma rays shielding. Hence, Bismuth based glasses can be better substitute to concrete walls at nuclear reactor sites and nuclear labs.

Heavy metal oxide glasses as gamma rays shielding material

Energy Technology Data Exchange (ETDEWEB)

Kaur, Preet; Singh, Devinder; Singh, Tejbir, E-mail: dr.tejbir@gmail.com

2016-10-15

The gamma rays shielding parameters for heavy metal oxide glasses and concrete samples are comparable. However, the transparent nature of glasses provides additional feature to visualize inside the shielding material. Hence, different researchers had contributed in computing/measuring different shielding parameters for different configurations of heavy metal oxide glass systems. In the present work, a detailed study on different heavy metal ({sub 56}Ba, {sub 64}Gd, {sub 82}Pb, {sub 83}Bi) oxide glasses has been presented on the basis of different gamma rays shielding parameters as reported by different researchers in the recent years. It has been observed that among the selected heavy metal oxide glass systems, Bismuth based glasses provide better gamma rays shielding. Hence, Bismuth based glasses can be better substitute to concrete walls at nuclear reactor sites and nuclear labs.

Effect of low dose electron beam irradiation on the alteration layer formed during nuclear glass leaching

Energy Technology Data Exchange (ETDEWEB)

Mougnaud, S., E-mail: sarah.mougnaud@gmail.com [CEA Marcoule, DEN, DTCD, SECM, BP 17171, 30207 Bagnols-sur-Cèze cedex (France); Tribet, M. [CEA Marcoule, DEN, DTCD, SECM, BP 17171, 30207 Bagnols-sur-Cèze cedex (France); Renault, J.-P. [NIMBE, CNRS, CEA, Université Paris Saclay, CEA Saclay, 91191 Gif-sur-Yvette cedex (France); Jollivet, P. [CEA Marcoule, DEN, DTCD, SECM, BP 17171, 30207 Bagnols-sur-Cèze cedex (France); Panczer, G. [Institut Lumière Matière, UMR 5306, Université Lyon 1-CNRS, Université de Lyon, 69622 Villeurbanne cedex (France); Charpentier, T. [NIMBE, CNRS, CEA, Université Paris Saclay, CEA Saclay, 91191 Gif-sur-Yvette cedex (France); Jégou, C. [CEA Marcoule, DEN, DTCD, SECM, BP 17171, 30207 Bagnols-sur-Cèze cedex (France)

2016-12-15

repolymerizes. - Highlights: • The nuclear glass/water interface is examined in the residual rate regime. • Leaching behavior of externally irradiated glass with electrons is studied. • Previous irradiation of the glass does not modify the alteration layer structure. • Evolution of alteration layer under external electronic irradiation is investigated. • Alteration layer is less sensitive to radiation induced defects creation than glass.

Aqueous corrosion of borosilicate glasses. Nature and properties of alteration layers

International Nuclear Information System (INIS)

Trotignon, Laurent

1990-01-01

This research thesis addresses physical and chemical processes which occur during aqueous corrosion of silicates, and the study of the properties of their interfaces with solutions, and thus issues related to the fate of high activity nuclear wastes which are embedded in a vitreous matrix as the potential release of radionuclides towards the environment then depends on the glass parcel behaviour submitted to chemical attacks which could alter it, notably by aqueous corrosion. The objective is then to model the dissolution of nuclear glass over long periods of time, and to predict the behaviour of radionuclides. The author compared the corrosion and alteration layers of gradually more complex borosilicate glasses, from a ternary sodium borosilicate glass to a simulated nuclear glass (the French reference glass R7T7). Complexity is increased by adding oxides. After some theoretical recalls on the structure and corrosion of borosilicate glasses, the author presents the studied materials, the corrosion experiments, and analytical techniques used to study alteration layers. The mechanism of formation of altered layers is studied based on corrosion experiments performed at 90 C on the whole set of glasses. Alteration layers formed on corroded glasses are studied and compared by using various techniques: electronic microscopy, high energy ion beams, spectroscopy, infrared, photo-electron spectroscopy. Implications for underground storage of nuclear glasses are discussed

Silicon isotope ratio measurements by inductively coupled plasma tandem mass spectrometry for alteration studies of nuclear waste glasses.

Science.gov (United States)

Gourgiotis, Alkiviadis; Ducasse, Thomas; Barker, Evelyne; Jollivet, Patrick; Gin, Stéphane; Bassot, Sylvain; Cazala, Charlotte

2017-02-15

High-level, long-lived nuclear waste arising from spent fuel reprocessing is vitrified in silicate glasses for final disposal in deep geologic formations. In order to better understand the mechanisms driving glass dissolution, glass alteration studies, based on silicon isotope ratio monitoring of 29 Si-doped aqueous solutions, were carried out in laboratories. This work explores the capabilities of the new type of quadrupole-based ICP-MS, the Agilent 8800 tandem quadrupole ICP-MS/MS, for accurate silicon isotope ratio determination for alteration studies of nuclear waste glasses. In order to avoid silicon polyatomic interferences, a new analytical method was developed using O 2 as the reaction gas in the Octopole Reaction System (ORS), and silicon isotopes were measured in mass-shift mode. A careful analysis of the potential polyatomic interferences on SiO + and SiO 2 + ion species was performed, and we found that SiO + ion species suffer from important polyatomic interferences coming from the matrix of sample and standard solutions (0.5M HNO 3 ). For SiO 2 + , no interferences were detected, and thus, these ion species were chosen for silicon isotope ratio determination. A number of key settings for accurate isotope ratio analysis like, detector dead time, integration time, number of sweeps, wait time offset, memory blank and instrumental mass fractionation, were considered and optimized. Particular attention was paid to the optimization of abundance sensitivity of the quadrupole mass filter before the ORS. We showed that poor abundance sensitivity leads to a significant shift of the data away from the Exponential Mass Fractionation Law (EMFL) due to the spectral overlaps of silicon isotopes combined with different oxygen isotopes (i.e. 28 Si 16 O 18 O + , 30 Si 16 O 16 O + ). The developed method was validated by measuring a series of reference solutions with different 29 Si enrichment. Isotope ratio trueness, uncertainty and repeatability were found to be

A simple method for tuning the glass transition process in inorganic phosphate glasses

Science.gov (United States)

Fulchiron, René; Belyamani, Imane; Otaigbe, Joshua U.; Bounor-Legaré, Véronique

2015-02-01

The physical modification of glass transition temperature (Tg) and properties of materials via blending is a common practice in industry and academia and has a large economic advantage. In this context, simple production of hitherto unattainable new inorganic glass blends from already existing glass compositions via blending raises much hope with the potential to provide new glasses with new and improved properties, that cannot be achieved with classical glass synthesis, for a plethora of applications such as computers screens, glass-to-metal seals, and storage materials for nuclear wastes. Here, we demonstrate that blends of the specific glass compositions studied are miscible in all proportions, an unreported phenomenon in hard condensed matter like glass. Interestingly, excellent agreement was found between the obtained data and calculated Tgs from theoretical equations (Supplementary information) for predicting the composition dependence of Tg for miscible blends with weak but significant specific interactions between the blend components. That this blending method is at present not applied to inorganic glasses reflects the fact that water and chemically resistant phosphate glasses with relatively low Tgs have become available only recently.

Iron phosphate glass containing simulated fast reactor waste: Characterization and comparison with pristine iron phosphate glass

International Nuclear Information System (INIS)

Joseph, Kitheri; Asuvathraman, R.; Venkata Krishnan, R.; Ravindran, T.R.; Govindaraj, R.; Govindan Kutty, K.V.; Vasudeva Rao, P.R.

2014-01-01

Detailed characterization was carried out on an iron phosphate glass waste form containing 20 wt.% of a simulated nuclear waste. High temperature viscosity measurement was carried out by the rotating spindle method. The Fe 3+ /Fe ratio and structure of this waste loaded iron phosphate glass was investigated using Mössbauer and Raman spectroscopy respectively. Specific heat measurement was carried out in the temperature range of 300–700 K using differential scanning calorimeter. Isoconversional kinetic analysis was employed to understand the crystallization behavior of the waste loaded iron phosphate glass. The glass forming ability and glass stability of the waste loaded glass were also evaluated. All the measured properties of the waste loaded glass were compared with the characteristics of pristine iron phosphate glass

Hydrolysis of R7T7 nuclear waste glass in dilute media: mechanisms and rate as a function of pH

International Nuclear Information System (INIS)

Advocat, T.; Vernaz, E.; Charpentier, H.; Crovisier, J.L.; Ehret, G.

1990-01-01

R7T7 nuclear waste glass dissolution in highly dilute aqueous media under static conditions at 90 0 C occurs according to two different mechanisms depending on the solution acidity. In acid media (pH 4.8 and 5.5), preferential extraction of glass network modifiers results in the formation of an alkali metal-depleted surface region on which amorphous and crystallized (phosphate) compounds rich in transition elements precipitate. Steady-state dissolution conditions are not reached, as attested by variable normalized Si, B and Na mass losses. Glass dissolution is stoichiometric in basic media (pH 7 to 10): the strong bonds of the silicated network are broken at a rate that increases with the pH: the glass dissolution rate increases by a factor of 15 between pH 7 and 10. Under these conditions, alteration products at the glass/solution interface do not constitute a short-term kinetic barrier against the release of the major glass components

Reference document on the long life behavior of nuclear glasses

International Nuclear Information System (INIS)

Godon, N.

2004-01-01

This document exposes the scientific analysis of the operational modelizations concerning the behavior of glasses (C wastes) for the long time storage and for the retrieval or ultimate underground disposal. The scientific approach adopted to establish the behavior of glasses uses a methodology, a strategy and defined approaches, described in this document. The containment glasses specifications, the glass behavior dry or in non saturated open environment, the glass behavior in aqueous environment, predictions models of glasses alteration and elements of validation are also presented. (A.L.B.)

Use of depleted uranium silicate glass to minimize release of radionuclides from spent nuclear fuel waste packages

International Nuclear Information System (INIS)

Forsberg, C.W.

1996-01-01

A Depleted Uranium Silicate Container Backfill System (DUSCOBS) is proposed that would use small, isotopically-depleted uranium silicate glass beads as a backfill material inside repository waste packages containing spent nuclear fuel (SNF). The uranium silicate glass beads would fill the void space inside the package including the coolant channels inside SNF assemblies. Based on preliminary analysis, the following benefits have been identified. DUSCOBS improves repository waste package performance by three mechanisms. First, it reduces the radionuclide releases from SNF when water enters the waste package by creating a local uranium silicate saturated groundwater environment that suppresses (a) the dissolution and/or transformation of uranium dioxide fuel pellets and, hence, (b) the release of radionuclides incorporated into the SNF pellets. Second, the potential for long-term nuclear criticality is reduced by isotopic exchange of enriched uranium in SNF with the depleted uranium (DU) in the glass. Third, the backfill reduces radiation interactions between SNF and the local environment (package and local geology) and thus reduces generation of hydrogen, acids, and other chemicals that degrade the waste package system. Finally, DUSCOBS provides a potential method to dispose of significant quantities of excess DU from uranium enrichment plants at potential economic savings. DUSCOBS is a new concept. Consequently, the concept has not been optimized or demonstrated in laboratory experiments

Shattered glass seeking the densest matter: the color glass condensate

CERN Multimedia

Appell, D

2004-01-01

"Physicists investigating heavy-particle collisions believe they are on the track of a universal form of matter, one common to very high energy particles ranging from protons to heavy nuclei such as uranium. Some think that this matter, called a color glass condensate, may explain new nuclear properties and the process of particle formation during collisions. Experimentalists have recently reported intriguing data that suggest a color glass condensate has actually formed in past work" (1 page)

Production and characterization of red mud based on glasses for the immobilization of nuclear wastes

International Nuclear Information System (INIS)

Vieira, Heveline

2015-01-01

Glasses based on red mud, a residual material from bauxite processing, were developed and characterized in this work. In order to promote its use, a minimum 60 wt% of red mud was used in the production of the glasses. According to XRD results, materials containing considerable amorphous phases were produced when using red mud as raw material. These amorphous phases were observed even though crystalline phases associated to Fe coming from the red mud itself were present. The material denominated 60L40S, which has a nominal composition of 60 wt% red mud showed the best properties comparing with the others compositions studied. However, these materials presented a high melting temperature. Changes in the composition of this material were made with the objective of lowering this temperature. Results indicated that the changes made to the material were successful in the reduction of the melting temperature. However, a reduction in the chemical properties of the resulting material was observed. Elements usually found in the chemical composition of nuclear wastes were added to the glasses produced. It was done with the objective of determining the effect of these elements on the chemical and physical properties of the red mud based glasses obtained. It was found that it was possible to add up to 15 wt% of these elements to the materials produced. The addition of these simulant materials promoted a reduction in the melting temperature of the resulting material. Above 15 wt%, the added elements precipitate in the structure of the resulting material. Even though the reduction in the chemical durability of the 60L40S material when simulant elements were added, it was observed that this material contained the simulant elements confined in its structure when in contact with water. This is a promising result, since it indicates that the 60L40S has the potential to immobilize elements from nuclear wastes . (author)

Structural and crystallisation study of a rare earth alumino borosilicate glass designed for nuclear waste confinement

International Nuclear Information System (INIS)

Quintas, A.

2007-09-01

This work is devoted to the study of a rare earth alumino borosilicate glass, which molar composition is 61,81 SiO 2 - 3,05 Al 2 O 3 - 8,94 B 2 O 3 - 14,41 Na 2 O - 6,33 CaO - 1,90 ZrO 2 - 3,56 Nd 2 O 3 , and envisaged for the immobilization of nuclear wastes originating from the reprocessing of high discharge burn up spent fuel. From a structural viewpoint, we investigated the role of the modifier cations on the arrangement of the glass network through different modifications of the glass composition: variation of the Na/Ca ratio and modification of the nature of the alkali and alkaline earth cations. The NMR and Raman spectroscopic techniques were useful to determine the distribution of modifier cations among the glass network and also to cast light on the competition phenomena occurring between alkali and alkaline earth cations for charge compensation of [AlO 4 ] - and [BO 4 ] - species. The neodymium local environment could be probed by optical absorption and EXAFS spectroscopies which enabled to better understand the insertion mode of Nd 3+ ions among the silicate domains of the glass network. Concerning the crystallization behavior we were interested in how the glass composition may influence the crystallization processes and especially the formation of the apatite phase of composition Ca 2 Nd 8 (SiO 4 ) 6 O 2 . In particular, this work underlined the important role of both alkaline earth and rare earth cations on the crystallization of the apatite phase. (author)

Volatility mechanisms of borosilicate glasses and molten glasses of nuclear interest structural effects

International Nuclear Information System (INIS)

Delorme, L.

1998-01-01

This work is devoted to the study of the mechanisms which control the volatility of the reference glass used for the confinement of radioactive waste. It was conducted on simplified compositions, in the SiO 2 -B 2 O 3 -Al 2 O 3 -αNa 2 O-(1-alpha)Li 2 O-CaO system.The structural approach carried out by NMR, from room temperature up to 1500 deg.C, shows a strong increase in the mobility of alkalis above Tg. A rapid exchange between B III and B IV sites near 700 deg.C, and the change of coordination number B IV- B III near 1100 deg.C, also seem to take place. The analysis of the vapor phase, carried out by High Temperature Mass Spectrometry coupled to Knudsen cells, reveals the presence between 780 deg.C and 830 deg.C of NaBO 2 (g), LiBO 2 (g) and Na 2 (BO 2 )2(g). The calculation of the partial pressure of each species shows that the total pressure of simplified glasses is dominated by the contribution of sodium. To study the volatility of glasses at higher temperature, equipment using the Transpiration method was used. The analysis of the deposits indicate the presence at 1060 deg.C of the species quoted previously. The vaporization rate and the vapor density were determined for each composition studied in a saturated state. Thus, we show that the volatility of the reference glass can be simulated by that of a simplified glass. For α=1, the kinetic of vaporization between 1060 deg.C and 1200 deg.C reveals an evaporation from the surface associated with a mechanism of diffusion in the molten glass. This is similar to the volatility of the reference glass at 1060 deg.C. To finally explain these mechanisms on a microscopic basis, we develop a model of molecular interactions. Between 780 deg.C and 830 deg.C, these mechanisms are controlled by a strong attraction between Na 2 O and Li 2 O, which maintains the total vapor pressure on a quasi-constant lever up to α=0.27. (author)

Plutonium immobilization in glass and ceramics

Energy Technology Data Exchange (ETDEWEB)

Knecht, D.A. [Lockheed Martin Idaho Technologies, Idaho Falls (United States); Murphy, W.M. [Southwest Research Institute, San Antonio, TX (United States)

1996-05-01

The Materials Research Society Nineteenth Annual Symposium on the Scientific Basis for Nuclear Waste Management was held in Boston on November 27 to December 1, 1995. Over 150 papers were presented at the Symposium dealing with all aspects of nuclear waste management and disposal. Fourteen oral sessions and on poster session included a Plenary session on surplus plutonium dispositioning and waste forms. The proceedings, to be published in April, 1996, will provide a highly respected, referred compilation of the state of scientific development in the field of nuclear waste management. This paper provides a brief overview of the selected Symposium papers that are applicable to plutonium immobilization and plutonium waste form performance. Waste forms that were described at the Symposium cover most of the candidate Pu immobilization options under consideration, including borosilicate glass with a melting temperature of 1150 {degrees}C, a higher temperature (1450 {degrees}C) lanthanide glass, single phase ceramics, multi-phase ceramics, and multi-phase crystal-glass composites (glass-ceramics or slags). These Symposium papers selected for this overview provide the current status of the technology in these areas and give references to the relevant literature.

Plutonium immobilization in glass and ceramics

International Nuclear Information System (INIS)

Knecht, D.A.; Murphy, W.M.

1996-01-01

The Materials Research Society Nineteenth Annual Symposium on the Scientific Basis for Nuclear Waste Management was held in Boston on November 27 to December 1, 1995. Over 150 papers were presented at the Symposium dealing with all aspects of nuclear waste management and disposal. Fourteen oral sessions and on poster session included a Plenary session on surplus plutonium dispositioning and waste forms. The proceedings, to be published in April, 1996, will provide a highly respected, referred compilation of the state of scientific development in the field of nuclear waste management. This paper provides a brief overview of the selected Symposium papers that are applicable to plutonium immobilization and plutonium waste form performance. Waste forms that were described at the Symposium cover most of the candidate Pu immobilization options under consideration, including borosilicate glass with a melting temperature of 1150 degrees C, a higher temperature (1450 degrees C) lanthanide glass, single phase ceramics, multi-phase ceramics, and multi-phase crystal-glass composites (glass-ceramics or slags). These Symposium papers selected for this overview provide the current status of the technology in these areas and give references to the relevant literature

Predicting liquid immiscibility in multicomponent nuclear waste glasses

International Nuclear Information System (INIS)

Peeler, D.K.; Hrma, P.R.

1994-04-01

Taylor's model for predicting amorphous phase separation in complex, multicomponent systems has been applied to high-level (simulated) radioactive waste glasses at the US Department of Energy's Hanford site. Taylor's model is primarily based on additions of modifying cations to a Na 2 O-B 2 O 3 -SiO 2 (NBS) submixture of the multicomponent glass. The position of the submixture relative to the miscibility dome defines the development probability of amorphous phase separation. Although prediction of amorphous phase separation in Hanford glasses (via experimental SEM/TEM analysis) is the primary thrust of this work; reported durability data is also provides limited insight into the composition/durability relationship. Using a modified model similar to Taylor's, the results indicate that immiscibility may be predicted for multicomponent waste glasses by the addition of Li 2 O to the ''alkali'' corner of the NBS submixture

Properties and solubility of chrome in iron alumina phosphate glasses containing high level nuclear waste

International Nuclear Information System (INIS)

Huang, W.; Day, D.E.; Ray, C.S.; Kim, C.W.; Reis, S.T.D.

2004-01-01

Chemical durability, glass formation tendency, and other properties of iron alumina phosphate glasses containing 70 wt% of a simulated high level nuclear waste (HLW), doped with different amounts of Cr 2 O 3 , have been investigated. All of the iron alumina phosphate glasses had an outstanding chemical durability as measured by their small dissolution rate (1 . 10 -9 g/(cm 2 . min)) in deionized water at 90 C for 128 d, their low normalized mass release as determined by the product consistency test (PCT) and a barely measurable corrosion rate of 2 . d) after 7 d at 200 C by the vapor hydration test (VHT). The solubility limit for Cr 2 O 3 in the iron phosphate melts was estimated at 4.1 wt%, but all of the as-annealed melts contained a few percent of crystalline Cr 2 O 3 that had no apparent effect on the chemical durability. The chemical durability was unchanged after deliberate crystallization, 48 h at 650 C. These iron phosphate waste forms, with a waste loading of at least 70 wt%, can be readily melted in commercial refractory crucibles at 1250 C for 2 to 4 h, are resistant to crystallization, meet all current US Department of Energy requirements for chemical durability, and have a solubility limit for Cr 2 O 3 which is at least three times larger than that for borosilicate glasses. (orig.)

Kinetics of Cold-Cap Reactions for Vitrification of Nuclear Waste Glass Based on Simultaneous Differential Scanning Calorimetry - Thermogravimetry (DSC-TGA) and Evolved Gas Analysis (EGA)

Energy Technology Data Exchange (ETDEWEB)

Rodriguez, Carmen P.; Pierce, David A.; Schweiger, Michael J.; Kruger, Albert A.; Chun, Jaehun; Hrma, Pavel R.

2013-12-03

For vitrifying nuclear waste glass, the feed, a mixture of waste with glass-forming and modifying additives, is charged onto the cold cap that covers 90-100% of the melt surface. The cold cap consists of a layer of reacting molten glass floating on the surface of the melt in an all-electric, continuous glass melter. As the feed moves through the cold cap, it undergoes chemical reactions and phase transitions through which it is converted to molten glass that moves from the cold cap into the melt pool. The process involves a series of reactions that generate multiple gases and subsequent mass loss and foaming significantly influence the mass and heat transfers. The rate of glass melting, which is greatly influenced by mass and heat transfers, affects the vitrification process and the efficiency of the immobilization of nuclear waste. We studied the cold-cap reactions of a representative waste glass feed using both the simultaneous differential scanning calorimetry thermogravimetry (DSC-TGA) and the thermogravimetry coupled with gas chromatography-mass spectrometer (TGA-GC-MS) as complementary tools to perform evolved gas analysis (EGA). Analyses from DSC-TGA and EGA on the cold-cap reactions provide a key element for the development of an advanced cold-cap model. It also helps to formulate melter feeds for higher production rate.

Overview of chemical modeling of nuclear waste glass dissolution

International Nuclear Information System (INIS)

Bourcier, W.L.

1991-02-01

Glass dissolution takes place through metal leaching and hydration of the glass surface accompanied by development of alternation layers of varying crystallinity. The reaction which controls the long-term glass dissolution rate appears to be surface layer dissolution. This reaction is reversible because the buildup of dissolved species in solution slows the dissolution rate due to a decreased dissolution affinity. Glass dissolution rates are therefore highly dependent on silica concentrations in solution because silica is the major component of the alteration layer. Chemical modeling of glass dissolution using reaction path computer codes has successfully been applied to short term experimental tests and used to predict long-term repository performance. Current problems and limitations of the models include a poorly defined long-term glass dissolution mechanism, the use of model parameters determined from the same experiments that the model is used to predict, and the lack of sufficient validation of key assumptions in the modeling approach. Work is in progress that addresses these issues. 41 refs., 7 figs., 2 tabs

Predicting liquid immiscibility in multicomponent nuclear waste glasses

International Nuclear Information System (INIS)

Peeler, D.K.; Hrma, P.R.

1994-01-01

Taylor's model for predicting amorphous phase separation in complex, multicomponent systems has been applied to high-level (simulated) radioactive waste glasses at the U.S. Department of Energy's Hanford site. Taylor's model is primarily based on additions of modifying cations to a Na 2 O-B 2 O 3 -SiO 2 (NBS) submixture of the multicomponent glass. The position of the submixture relative to the immiscibility dome defines the development probability of amorphous phase separation. Although prediction of amorphous phase separation in Hanford glasses (via experimental SEM/TEM analysis) is the primary thrust of this work; reported durability data is also provides limited insight into the composition/durability relationship. Using a modified model similar to Taylor's, the results indicate that immiscibility may be predicted for multicomponent waste glasses by the addition of Li 2 O to the open-quotes alkaliclose quotes corner of the NBS submixture

Irradiations effects on the structure of boro-silicated glasses: long term behaviour of nuclear waste glassy matrices

International Nuclear Information System (INIS)

Bonfils, J. de

2007-09-01

This work deals with the long term behaviour of R7T7-type nuclear waste glasses and more particularly of non-active boro-silicated glasses made up of 3 or 5 oxides. Radioactivity of active glasses is simulated by multi energies ions implantations which reproduce the same defects. The damages due to the alpha particles are simulated by helium ions implantations and those corresponding to the recoil nucleus are obtained with gold ions ones. Minor actinides, stemming from the used fuel, is simulated by trivalent rare-earths (Eu 3+ and Nd 3+ ). In a first part, we have shown by macroscopic experiments (Vickers hardness - swelling) and optical spectroscopies (Raman - ATR-IR) that the structure of the glassy matrices is modified under implantations until a dose of 2,3.10 13 at.cm -2 , which corresponds to a R7T7 storage time estimated at 300 years. Beyond this dose, no additional modifications have been observed. The second part concerns the local environment of the rare-earth ions in glasses. Two different environments were found and identified as follows: one is a silicate rich one and the other is attributed to a borate rich one. (author)

Isothermal crystallization kinetics in simulated high-level nuclear waste glass

International Nuclear Information System (INIS)

Vienna, J.D.; Hrma, P.; Smith, D.E.

1997-01-01

Crystallization kinetics of a simulated high-level waste (HLW) glass were measured and modelled. Kinetics of acmite growth in the standard HW39-4 glass were measured using the isothermal method. A time-temperature-transformation (TTT) diagram was generated from these data. Classical glass-crystal transformation kinetic models were empirically applied to the crystallization data. These models adequately describe the kinetics of crystallization in complex HLW glasses (i.e., RSquared = 0.908). An approach to measurement, fitting, and use of TTT diagrams for prediction of crystallinity in a HLW glass canister is proposed

Silicon isotope ratio measurements by inductively coupled plasma tandem mass spectrometry for alteration studies of nuclear waste glasses

Energy Technology Data Exchange (ETDEWEB)

Gourgiotis, Alkiviadis, E-mail: alkiviadis.gourgiotis@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PRP-DGE/SRTG/LT2S, Fontenay-aux-Roses (France); Ducasse, Thomas [CEA, DEN, DTCD, SECM, F-30207 Bagnols-sur-Cèze (France); Barker, Evelyne [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PRP-DGE/SRTG/LT2S, Fontenay-aux-Roses (France); Jollivet, Patrick; Gin, Stéphane [CEA, DEN, DTCD, SECM, F-30207 Bagnols-sur-Cèze (France); Bassot, Sylvain; Cazala, Charlotte [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PRP-DGE/SRTG/LT2S, Fontenay-aux-Roses (France)

2017-02-15

High-level, long-lived nuclear waste arising from spent fuel reprocessing is vitrified in silicate glasses for final disposal in deep geologic formations. In order to better understand the mechanisms driving glass dissolution, glass alteration studies, based on silicon isotope ratio monitoring of {sup 29}Si-doped aqueous solutions, were carried out in laboratories. This work explores the capabilities of the new type of quadrupole-based ICP-MS, the Agilent 8800 tandem quadrupole ICP-MS/MS, for accurate silicon isotope ratio determination for alteration studies of nuclear waste glasses. In order to avoid silicon polyatomic interferences, a new analytical method was developed using O{sub 2} as the reaction gas in the Octopole Reaction System (ORS), and silicon isotopes were measured in mass-shift mode. A careful analysis of the potential polyatomic interferences on SiO{sup +} and SiO{sub 2}{sup +} ion species was performed, and we found that SiO{sup +} ion species suffer from important polyatomic interferences coming from the matrix of sample and standard solutions (0.5M HNO{sub 3}). For SiO{sub 2}{sup +}, no interferences were detected, and thus, these ion species were chosen for silicon isotope ratio determination. A number of key settings for accurate isotope ratio analysis like, detector dead time, integration time, number of sweeps, wait time offset, memory blank and instrumental mass fractionation, were considered and optimized. Particular attention was paid to the optimization of abundance sensitivity of the quadrupole mass filter before the ORS. We showed that poor abundance sensitivity leads to a significant shift of the data away from the Exponential Mass Fractionation Law (EMFL) due to the spectral overlaps of silicon isotopes combined with different oxygen isotopes (i.e. {sup 28}Si{sup 16}O{sup 18}O{sup +}, {sup 30}Si{sup 16}O{sup 16}O{sup +}). The developed method was validated by measuring a series of reference solutions with different {sup 29}Si

Composition and redox control of waste glasses: Recommendation for process control limit

International Nuclear Information System (INIS)

Jantzen, C.M.; Plodinec, M.J.

1986-01-01

An electrochemical series of redox couples, originally developed for Savannah River Laboratory glass frit 131 (SRL-131) as a reference composition, has been extended to two other alkali borosilicate compositions that are candidate glasses for nuclear waste immobilization. Since no dramatic differences were ascertained in the redox chemistry of selected multivalent elements in SRL-131 versus that in Savannah River Laboratory glass frit 165 (SRL-165) and in West Valley glass number-sign 205 (WV-205), the comprehensive electrochemical series can readily be applied to a range of nuclear waste glass compositions. In order to alleviate potential problems with foaming and precipitation of insolubles during the processing of the nuclear waste in these glass melts, the [Fe 2+ ]/[Fe 3+ ] ratio of the melt should be between 0.1 and 0.5. 27 refs., 4 figs., 2 tabs

Thermodynamical modeling of nuclear glasses: coexistence of amorphous phases; Modelisation thermodynamique des verres nucleaires: coexistence entre phases amorphes

Energy Technology Data Exchange (ETDEWEB)

Adjanor, G

2007-11-15

Investigating the stability of borosilicate glasses used in the nuclear industry with respect to phase separation requires to estimate the Gibbs free energies of the various phases appearing in the material. In simulation, using current computational resources, a direct state-sampling of a glassy system with respect to its ensemble statistics is not ergodic and the estimated ensemble averages are not reliable. Our approach consists in generating, at a given cooling rate, a series of quenches, or paths connecting states of the liquid to states of the glass, and then in taking into account the probability to generate the paths leading to the different glassy states in ensembles averages. In this way, we introduce a path ensemble formalism and calculate a Landau free energy associated to a glassy meta-basin. This method was validated by accurately mapping the free energy landscape of a 38-atom glassy cluster. We then applied this approach to the calculation of the Gibbs free energies of binary amorphous Lennard-Jones alloys, and checked the correlation between the observed tendencies to order or to phase separate and the computed Gibbs free energies. We finally computed the driving force to phase separation in a simplified three-oxide nuclear glass modeled by a Born-Mayer-Huggins potential that includes a three-body term, and we compared the estimated quantities to the available experimental data. (author)

Development Of Glass Matrices For HLW Radioactive Wastes

International Nuclear Information System (INIS)

Jantzen, C.

2010-01-01

Vitrification is currently the most widely used technology for the treatment of high level radioactive wastes (HLW) throughout the world. Most of the nations that have generated HLW are immobilizing in either borosilicate glass or phosphate glass. One of the primary reasons that glass has become the most widely used immobilization media is the relative simplicity of the vitrification process, e.g. melt waste plus glass forming frit additives and cast. A second reason that glass has become widely used for HLW is that the short range order (SRO) and medium range order (MRO) found in glass atomistically bonds the radionuclides and governs the melt properties such as viscosity, resistivity, sulphate solubility. The molecular structure of glass controls contaminant/radionuclide release by establishing the distribution of ion exchange sites, hydrolysis sites, and the access of water to those sites. The molecular structure is flexible and hence accounts for the flexibility of glass formulations to waste variability. Nuclear waste glasses melt between 1050-1150 C which minimizes the volatility of radioactive components such as Tc 99 , Cs 137 , and I 129 . Nuclear waste glasses have good long term stability including irradiation resistance. Process control models based on the molecular structure of glass have been mechanistically derived and have been demonstrated to be accurate enough to control the world's largest HLW Joule heated ceramic melter in the US since 1996 at 95% confidence.

Evaluation of lead-iron-phosphate glass as a high-level waste form

International Nuclear Information System (INIS)

Chick, L.A.; Bunnell, L.R.; Strachan, D.M.; Kissinger, H.E.; Hodges, F.N.

1986-09-01

The lead-iron-phosphate (Pb-Fe-P) glass developed at Oak Ridge National Laboratory was evaluated for its potential as an improvement over the current reference nuclear waste form, borosilicate (B-Si) glass. The evaluation was conducted as part of the Second Generation HLW Technology Subtask of the Nuclear Waste Treatment Program at Pacific Northwest Laboratory. The purpose of this work was to investigate possible alternatives to B-Si glass as second-generation waste forms. While vitreous Pb-Fe-P glass appears to have substantially better chemical durability than B-Si glass, severe crystallization or devitrification leading to deteriorated chemical durability would result if this glass were poured into large canisters as is the procedure with B-Si glass. Cesium leach rates from this crystallized material are orders of magnitude greater than those from B-Si glass. Therefore, to realize the potential performance advantages of the Pb-Fe-P material in a nuclear waste form, the processing method would have to cool the material rapidly to retain its vitreous structure

Relative leach behavior of waste glasses and naturally occurring glasses

International Nuclear Information System (INIS)

Adams, P.B.

1979-01-01

Simulated nuclear waste glasses of the sodium-borosilicate type with a low waste loading and of the zinc-borosilicate type with a high waste loading have been compared with obsidians. The resuls indicate that the waste glasses would corrode in normal natural environments at a rate of about 0.1 μm per year at 30 0 C and about 5 μm per year at 90 0 C, compared with obsidians which seem to corrode at, or less than, about 0.01 μm per year at 30 0 C and less than 1 μm per year at 90 0 C. Activation energies for reactions of the two waste glasses with pure water are about 20 kcal/g-mol. 3 figures, 7 tables

Durability of Silicate Glasses: An Historical Approach

International Nuclear Information System (INIS)

Farges, Francois; Etcheverry, Marie-Pierre; Haddi, Amine; Trocellier, Patrick; Curti, Enzo; Brown, Gordon E. Jr.

2007-01-01

We present a short review of current theories of glass weathering, including glass dissolution, and hydrolysis of nuclear waste glasses, and leaching of historical glasses from an XAFS perspective. The results of various laboratory leaching experiments at different timescales (30 days to 12 years) are compared with results for historical glasses that were weathered by atmospheric gases and soil waters over 500 to 3000 years. Good agreement is found between laboratory experiments and slowly leached historical glasses, with a strong enrichment of metals at the water/gel interface. Depending on the nature of the transition elements originally dissolved in the melt, increasing elemental distributions are expected to increase with time for a given glass durability context

Durability of Silicate Glasses: An Historical Approach

Energy Technology Data Exchange (ETDEWEB)

Farges, Francois; /Museum Natl. Hist. Natur. /Stanford U., Geo. Environ. Sci.; Etcheverry, Marie-Pierre; /Marne la Vallee U.; Haddi, Amine; /Marne la Valle U.; Trocellier,; /Saclay; Curti, Enzo; /PSI, Villigen; Brown, Gordon E., Jr.; /SLAC, SSRL

2007-01-02

We present a short review of current theories of glass weathering, including glass dissolution, and hydrolysis of nuclear waste glasses, and leaching of historical glasses from an XAFS perspective. The results of various laboratory leaching experiments at different timescales (30 days to 12 years) are compared with results for historical glasses that were weathered by atmospheric gases and soil waters over 500 to 3000 years. Good agreement is found between laboratory experiments and slowly leached historical glasses, with a strong enrichment of metals at the water/gel interface. Depending on the nature of the transition elements originally dissolved in the melt, increasing elemental distributions are expected to increase with time for a given glass durability context.

Strontium chloroapatite based glass-ceramics composites for nuclear waste immobilisation

International Nuclear Information System (INIS)

Jena, Hrudananda; Maji, Binoy Kumar; Asuvathraman, R.; Govindan Kutty, K.V.

2013-01-01

Apatites are naturally occurring minerals with a general formula of M 10 (PO 4 ) 6 X 2 , (M= Ca, Sr, Ba, X= OH, Cl, F) with a hexagonal crystal structure (S.G :P6 3 /m) and can accommodate alkaline earth and various other aliovalent cations and anions into its crystal structure. Apatites are also known to have high resistance to leaching of the constituent elements under geological conditions. It may not often be possible to immobilize the whole spectrum of the radioactive waste in a single phase M 10 (PO 4 ) 6 Cl 2 , then a combination of M-chloroapatite encapsulated in borosilicate glass (BSG) can immobilize most of the radwaste elements in the composite glass-ceramic matrix (glass bonded chloroapatite), thus utilizing the immobilizing efficiency of both the ceramic phase and glass. In the present study, the synthesis, characterization and thermo-physical property measurements of the Sr-chloroapatite (SrApCI) and some glass-bonded composites based on it have been investigated. The Sr-chloroapatite glass-ceramics were prepared by solid state reactions among stoichiometric concentrations of apatite forming reagents, 20 wt. % borosilicate glass (BSG), and known concentrations (10, 13 and 16 wt. %) of a simulated waste in chloride form. The products were characterized by XRD to confirm the formation of Sr 10 (PO 4 ) 6 Cl 2 and glass bonded-chloroapatite composites. The surface morphology and qualitative chemical composition of the powders were examined by SEM and EDX. Thermal expansion and glass transition temperature of the matrices were measured by dilatometry. Glass transition temperature of the glass-bonded composites was also examined by differential scanning calorimetry and differential thermal analysis. The 10-16 wt.% waste loaded matrices showed similar thermal expansion as that of SrApCI, indicating the thermal stability of the matrix to chloride waste immobilization. The glass transition temperature of the waste loaded matrices decreases on increasing the

START - glass model of PWR

International Nuclear Information System (INIS)

Marn, J.; Ramsak, M.

1998-01-01

Recognizing the importance of nuclear engineering in the area of process engineering the University of Maribor, Faculty of Mechanical Engineering has invested in procuring and erecting glass model of pressurized water reactor. This paper deals with description of the model, its capabilities, and plans for its use within nuclear engineering community of Slovenia. The model, made primarily of glass, serves three purposes: educational, professional development and research. As an example, medium break loss of coolant accident is presented in the paper. Temperatures within primary and secondary side, and pressure on primary side of reactor coolant system are followed. The characteristic points are emphasized, and commented.(author)

Reaction of water with a simulated high-level nuclear waste glass at 3000C, 300 bars

International Nuclear Information System (INIS)

McCarthy, G.J.; Scheetz, B.E.; Komarneni, S.; Smith, D.K.

1978-01-01

The hydrothermal stability of high-level nuclear wastes is an important consideration in establishing waste form acceptance criteria for a geological repository in basalt. A detailed examination of the stability of a typical simulated high-level waste glass and pressurized water at 300 0 C in a closed system has shown that extensive reaction occurred within a few weeks. The water acted first as a catalyst-solvent in devitrification of the glass and in dissolution, transport, and recrystallization of some of its constituents, and, second, as a reactant in forming hydrated and hydroxylated phases. This reaction with water resulted in the conversion of a solid shard of glass into a fragmented and partially dispersed mass of crystalline and noncrystalline material plus dissolved species within two weeks. The major crystalline reaction products were found to be analogs of naturally occurring minerals: (Cs,Na,Rb) 2 (UO 2 ) 2 .(Si 2 O 5 ) 3 .4H 2 O (weeksite) and a series of pyroxene-structure phases, (Na,Ca) (Fe,Zn,Ti)Si 2 O 6 (acmite, acmite--augites). Weeksite, however, is not expected to have long-term stability in the basalt environment. Much of the Na and Mo, and almost all of the B, in the original glass was identified in the product solutions. Of the elements or analogs of long-lived, hazardous radionuclides studied in this work, only Cs was observed in these solutions in substantial amounts. Although the comparatively rapid and extensive reactions at 300 0 C would appear to require that an acceptable glass would have low waste and heat loading, it is suggested that there is good potential for favorable glass--basalt--water hydrothermal interactions. Favorable interactions would mean that, in the event of a hydrothermal incident, the interaction products would be more stable than the original waste form and would remain in the immediate repository

Glass leaching performance

International Nuclear Information System (INIS)

Chick, L.A.; Turcotte, R.P.

1983-05-01

Current understanding of the leaching performance of high-level nuclear waste (HLW) glass is summarized. The empirical model of waste glass leaching behavior developed shows that at high water flow rates the glass leach rate is kinetically limited to a maximum value. At intermediate water flow rates, leaching is limited by the solution concentration of silica and decreases with decreasing water flow rates. Release of soluble elements is controlled by silica dissolution because silica forms the binding network of the glass. At low water flow rates, mass loss rates reach values controlled by formation rates of alteration minerals, or by diffusion of dissolution products through essentially stagnant water. The parameters reviewed with respect to their quantifiable influence on leaching behavior include temperature, pH, leachant composition, glass composition, thermal history, and radiation. Of these, temperature is most important since the rate of mass loss approximately doubles with each 10 0 C increase in dilute solutions. The pH has small effects within the 4 to 10 range. The chemical composition of the leachant is most important with regard to its influence on alteration product formation. Glass composition exhibits the largest effects at high flow rates where improved glasses leach from ten to thirty times slower than glass 76 to 68. The effects of the thermal history (devitrification) of the glass are not likely to be significant. Radiation effects are important primarily in that radiolysis can potentially drive pH values to less than 4. Radiation damage to the glass causes insignificant changes in leaching performance

Thermo-chemistry of nuclear waste glasses: a new approach; Thermochimie des verres de stockage de dechets nucleaires: une nouvelle approche

Energy Technology Data Exchange (ETDEWEB)

Linard, Y.; Neuville, D.R.; Richet, P. [Institut de Physique du Globe, Lab. de Physique des Geomateriaux, URA CNRS 734, 75 - Paris (France)

1997-07-01

Understanding of the stability and weathering of glasses used for storing fission products is hampered by a general lack of basic thermochemical information. Models have been setup to predict Gibbs free energies of dissolution of glasses, but ascertaining their accuracy is made difficult by the very lack of reliable experimental data with which model results should be compared. As enthalpies of formation can in principle be determined from usual solution calorimetry experiments, the lack of Gibbs-free energy data for glasses mainly stems from the fact that, as disordered substances, glasses do not obey the third principle and have indeed large configurational entropies. These entropies can be determined from thermochemical measurements only when there exist a congruently melting crystalline compound with the same composition. Using available data, we have calculated the Gibbs-free energies of formation of a series of silicate glasses for which such a calorimetric determination is possible. With these results, we assess the predictions of Paul's model (1977) for calculating Gibbs-free energies of dissolution. As the complex compositions of the borosilicate glasses used for nuclear waste storage prevent determining configurational entropies by calorimetric methods, we point out how these can be determined instead from viscosity measurements. We finally discuss the implications of this approach for modeling of water-glass interactions. (authors)

DEVELOPMENT OF GLASS MATRICES FOR HLW RADIOACTIVE WASTES

Energy Technology Data Exchange (ETDEWEB)

Jantzen, C.

2010-03-18

Vitrification is currently the most widely used technology for the treatment of high level radioactive wastes (HLW) throughout the world. Most of the nations that have generated HLW are immobilizing in either borosilicate glass or phosphate glass. One of the primary reasons that glass has become the most widely used immobilization media is the relative simplicity of the vitrification process, e.g. melt waste plus glass forming frit additives and cast. A second reason that glass has become widely used for HLW is that the short range order (SRO) and medium range order (MRO) found in glass atomistically bonds the radionuclides and governs the melt properties such as viscosity, resistivity, sulphate solubility. The molecular structure of glass controls contaminant/radionuclide release by establishing the distribution of ion exchange sites, hydrolysis sites, and the access of water to those sites. The molecular structure is flexible and hence accounts for the flexibility of glass formulations to waste variability. Nuclear waste glasses melt between 1050-1150 C which minimizes the volatility of radioactive components such as Tc{sup 99}, Cs{sup 137}, and I{sup 129}. Nuclear waste glasses have good long term stability including irradiation resistance. Process control models based on the molecular structure of glass have been mechanistically derived and have been demonstrated to be accurate enough to control the world's largest HLW Joule heated ceramic melter in the US since 1996 at 95% confidence.

An evaluation of electric melter refractories for contact with glass used for the immobilisation of nuclear waste

International Nuclear Information System (INIS)

Hayward, P.J.; George, I.M.

1987-01-01

Corrosion tests have been performed on twelve candidate refractories in contact with borosilicate, titanosilicate, and aluminosilicate melts, in order to rank them for use in an all-electric melter for the production of waste form materials suitable for immobilising nuclear fuel recycle wastes. Viscosities and electrical conductivities of the melts have also been measured to enable optimum processing conditions to be determined. Of the materials tested, the choice of glass contact refractory for the Joule heated melting of the borosilicate and titanosilicate compositions is Monofrax K3 or SEPR 2161, in conjunction with tin oxide electrodes. The aluminosilicate glass waste form would require an alternative method of production (sol-gel processing, or sintering of a precursor frit), because of its high viscosity. (author)

Assessment of water/glass interactions in waste glass melter operation

International Nuclear Information System (INIS)

Postma, A.K.; Chapman, C.C.; Buelt, J.L.

1980-04-01

A study was made to assess the possibility of a vapor explosion in a liquid-fed glass melter and during off-standard conditions for other vitrification processes. The glass melter considered is one designed for the vitrification of high-level nuclear wastes and is comprised of a ceramic-lined cavity with electrodes for joule heating and processing equipment required to add feed and withdraw glass. Vapor explosions needed to be considered because experience in other industrial processes has shown that violent interactions can occur if a hot liquid is mixed with a cooler, vaporizable liquid. Available experimental evidence and theoretical analyses indicate that destructive glass/water interactions are low probability events, if they are possible at all. Under standard conditions, aspects of liquid-fed melter operation which work against explosive interactions include: (1) the aqueous feed is near its boiling point; (2) the feed contains high concentrations of suspended particles; (3) molten glass has high viscosity (greater than 20 poise); and (4) the glass solidifies before film boiling can collapse. While it was concluded that vapor explosions are not expected in a liquid-fed melter, available information does not allow them to be ruled out altogether. Several precautionary measures which are easily incorporated into melter operation procedures were identified and additional experiments were recommended

Nuclear Spin Lattice Relaxation and Conductivity Studies of the Non-Arrhenius Conductivity Behavior in Lithium Fast Ion Conducting Sulfide Glasses

Energy Technology Data Exchange (ETDEWEB)

Meyer, Benjamin Michael [Iowa State Univ., Ames, IA (United States)

2003-01-01

As time progresses, the world is using up more of the planet's natural resources. Without technological advances, the day will eventually arrive when these natural resources will no longer be sufficient to supply all of the energy needs. As a result, society is seeing a push for the development of alternative fuel sources such as wind power, solar power, fuel cells, and etc. These pursuits are even occurring in the state of Iowa with increasing social pressure to incorporate larger percentages of ethanol in gasoline. Consumers are increasingly demanding that energy sources be more powerful, more durable, and, ultimately, more cost efficient. Fast Ionic Conducting (FIC) glasses are a material that offers great potential for the development of new batteries and/or fuel cells to help inspire the energy density of battery power supplies. This dissertation probes the mechanisms by which ions conduct in these glasses. A variety of different experimental techniques give a better understanding of the interesting materials science taking place within these systems. This dissertation discusses Nuclear Magnetic Resonance (NMR) techniques performed on FIC glasses over the past few years. These NMR results have been complimented with other measurement techniques, primarily impedance spectroscopy, to develop models that describe the mechanisms by which ionic conduction takes place and the dependence of the ion dynamics on the local structure of the glass. The aim of these measurements was to probe the cause of a non-Arrhenius behavior of the conductivity which has been seen at high temperatures in the silver thio-borosilicate glasses. One aspect that will be addressed is if this behavior is unique to silver containing fast ion conducting glasses. more specifically, this study will determine if a non-Arrhenius correlation time, τ, can be observed in the Nuclear Spin Lattice Relaxation (NSLR) measurements. If so, then can this behavior be modeled with a new single

Glass-Graphite Composite Materials

International Nuclear Information System (INIS)

Mayzan, M.Z.H.; Lloyd, J.W.; Heath, P.G.; Stennett, M.C.; Hyatt, N.C.; Hand, R.J.

2016-01-01

A summary is presented of investigations into the potential of producing glass-composite materials for the immobilisation of graphite or other carbonaceous materials arising from nuclear power generation. The methods are primarily based on the production of base glasses which are subsequently sintered with powdered graphite or simulant TRISO particles. Consideration is also given to the direct preparation of glass-graphite composite materials using microwave technology. Production of dense composite wasteforms with TRISO particles was more successful than with powdered graphite, as wasteforms containing larger amounts of graphite were resistant to densification and the glasses tried did not penetrate the pores under the pressureless conditions used. Based on the results obtained it is concluded that the production of dense glassgraphite composite wasteforms will require the application of pressure. (author)

Contribution to the study of the effects of {alpha}-irradiation in nuclear glasses; Contribution a l'etude des effets de l'irradiation {alpha} sur les verres nucleaires

Energy Technology Data Exchange (ETDEWEB)

Abbas, A

2001-07-01

The main topic of this work is to characterise the effects of {alpha}-disintegration in nuclear waste glasses. Experimental and numerical approaches have been considered. The structure of the French nuclear waste glass (R7T7) has been simulated using four- and six-oxides simplified glasses which contain the main elements of the R7T7 glass: SiO{sub 2}, B{sub 2}O{sub 3}, Na{sub 2}O, ZrO{sub 2}, Al{sub 2}O{sub 3} and CaO. Four- and six-oxides glasses have been irradiated with 1 MeV-He{sup +} (ionisation) and 2.1 MeV-Kr{sup 3+} (ionisation and atomic collisions) ions in order to reproduce the effects of the {alpha}-particle and of the recoil nucleus emitted during {alpha}-disintegration of actinides, and also to differentiate electronic and ballistic effects. Irradiated glasses have been characterised using several techniques, which have been adapted to the peculiarities of our samples (isolated material, small irradiated depth). The results point out the salient role of sodium in the observed modifications: depth concentration profiles obtained with RBS show an accumulation of sodium at the irradiated surface. We found a apparent acceleration of sodium release in leaching experiments which confirm that point. Modifications observed in Raman spectra of irradiated glasses show an increase of the polymerisation (increase of Q{sub 3}/Q{sub 2} ratio) due to sodium migration. In simplified glasses we have found that the modifications of mechanical properties by external irradiations reproduce the modifications observed in actinide doped nuclear glass (decrease of hardness and increase of fracture toughness). At the same time, we performed Molecular Dynamics simulations of a six-oxides glass. We have shown that the surface modifies the glass structure down to a depth of 10 Angstrom: modification of depth concentration profiles, decrease of the atomic coordination number (A1, B and Si). During cascades, we found that atomic displacements are easier near the surface. This

Improving the Modeling of Hydrogen Solubility in Heavy Oil Cuts Using an Augmented Grayson Streed (AGS Approach Modélisation améliorée de la solubilité de l’hydrogène dans des coupes lourdes par l’approche de Grayson Streed Augmenté (GSA

Directory of Open Access Journals (Sweden)

Torres R.

2013-04-01

Full Text Available The Grayson Streed (GS method [Grayson H.G. and Streed C.W. (1963 6th World Petroleum Congress , Frankfurt am Main, Germany, 19-26 June, pp. 169-181] is often used by the industry for calculating hydrogen solubility in petroleum fluids. However, its accuracy becomes very bad when very heavy fluids are considered. An improvement is proposed in this work, based on a Flory-augmented activity coefficient model. Hydrogen solubilities in n-alkanes from n−C7 up to n−C36 have been investigated and a decreasing Henry constant with molecular weight is evidenced. The analysis of the Henry constant behaviour with molecular weight suggests a simple improvement to the model, using a Flory entropic contribution, thus keeping its predictive character. This improvement led to the necessity of refitting a number of fundamental hydrogen parameters. The resulting model behaves better for heavy components and for aromatics. The petroleum fractions evaluated with the Augmented Grayson-Streed (AGS model are taken from Cai et al. [Cai H.Y. et al. (2001 Fuel 80, 1055-1063] and Lin et al. [Lin H.M. et al. (1981 Ind. Eng. Chem. Process Des. Dev. 20, 2, 253-256]. The importance of the petroleum fluid characterization is stressed. A sensitivity analysis has shown that the solubility parameter has a much larger effect than the other parameters: great care must be taken at calculating that property. The predictions of hydrogen solubility in petroleum fractions and in coal liquids were improved compared with the Grayson Streed model, resulting in an Absolute Average Deviation (AAD of 30% for AGS model compared to 55% for Grayson-Streed model, in the range of 80-380°C and 6.3-258.9 bar. La méthode de Grayson Streed (GS [Grayson H.G. and Streed C.W. (1963 6th World Petroleum Congress, Frankfurt am Main, Germany, 19-26 June, pp. 169-181] est souvent préconisée dans l’industrie pour calculer la solubilité de l’hydrogène dans des coupes pétrolières. Il se fait

Evaluation of lead-iron-phosphate glass as a high-level waste form

International Nuclear Information System (INIS)

Chick, L.A.; Bunnell, L.R.; Strachan, D.M.; Kissinger, H.E.; Hodges, F.N.

1986-01-01

The lead-iron-phosphate (Pb-Fe-P) nuclear waste glass developed at Oak Ridge National Laboratory (ORNL) was evaluated for its potential as an improvement over the current reference waste form, borosilicate (B-Si) glass. Vitreous Pb-Fe-P glass appears to have substantially better chemical durability than B-Si glass. However, severe crystallization leading to deteriorated chemical durability would result if this glass were poured into large canisters, as is presently done with B-Si glass. Cesium leach rates from this crystallized material are orders of magnitude greater than those from B-Si glass. Therefore, to realize the performance advantages of the Pb-Fe-P material in a nuclear waste form, it would be necessary to process it so that it is cooled rapidly, thus retaining its vitreous structure

Evaluation of lead-iron-phosphate glass as a high-level waste form

International Nuclear Information System (INIS)

Chick, L.A.; Bunnell, L.R.; Strachan, D.M.; Kissinger, H.E.; Hodges, F.N.

1986-01-01

The lead-iron-phosphate nuclear waste glass developed at Oak Ridge National Laboratory (ORNL) was evaluated for its potential as an improvement over the current reference waste form, borosilicate glass. Vitreous lead-iron-phosphate glass appears to have substantially better chemical durability than borosilicate glass. However, severe crystallization leading to deteriorated chemical durability would result if this glass were poured into large canisters as is presently done with borosilicate glass. Cesium leach rates from this crystallized material are orders of magnitude greater than those from borosilicate glass. Therefore, in order to realize the performance advantages of the lead-iron-phosphate material in a nuclear waste form, it would be necessary to process it so that it is rapidly cooled, thus retaining its vitreous structure. 22 refs., 4 figs., 4 tabs

A comparative property investigation of lithium phosphate glass ...

Indian Academy of Sciences (India)

2017-08-16

Aug 16, 2017 ... However, MW processing of bulk glass is a relatively recent development and a ... candidates for nuclear waste immobilization [19]. Low refrac- ... one of the basic prototype glasses in solid-state electrolyte, because of its high ...

SON68 glass alteration enhanced by magnetite

Energy Technology Data Exchange (ETDEWEB)

Godon, Nicole; Gin, Stephane; Rebiscoul, Diane; Frugier, Pierre [CEA, DEN-Marcoule, F30207, Bagnols-sur-Ceze (France)

2013-07-01

This paper reports experimental and modeling results of SON68 glass / magnetite interactions while in contact with synthetic groundwater from a clay environment. It is shown that magnetite enhances glass alteration, first by the sorption of Si released from the glass onto magnetite surfaces, then by a second process that could be the precipitation of an iron silicate mineral or the transformation of magnetite into a more reactive phase like hematite or goethite. This study globally suggests a detrimental effect of magnetite on the long-term durability of nuclear glass in geological disposal conditions. (authors)

Contribution to the application of nuclear microprobe in geochemistry. Carbon and nitrogen microanalysis in glasses and minerals

International Nuclear Information System (INIS)

Mosbah, M.

1988-01-01

The morphological complexity of geological materials implies the use of microanalysis techniques utilization. Nuclear microprobe allows selective and no destructive light elements determination, through nuclear reactions. Nuclear microanalysis has been used to characterize carbon and nitrogen in volatile phase dissolved in magmatic samples. The application of some microanalysis techniques in geochemistry are discussed, nuclear microprobe theory and techniques are developed. Minerals, glasses and glassy inclusions are described, and more particularly, the interest of these investigations. Optimal conditions of carbon and nitrogen analysis ( 12 C(d.p) 13 C and 14 N(d,p) 15 N reaction respectively), as deuteron energy and observation angle are studied. A methodology has been established for this purpose. Several results are exposed: Punctual analysis, carbon concentration profile in depth surface scanning, surficial mapping in glassy inclusions. The carbon content interpretation in glassy inclusions measured conveniently for the first time agrees with data obtained through other techniques. In conclusion, degazing schedule improvements require more analysis. Perspective research axis are evocated [fr

On fluidization of borosilicate glasses in intense radiation fields - 16055

International Nuclear Information System (INIS)

Ojovan, Michael; Moebus, Guenter; Tsai, Jim; Cook, Stuart; Yang, Guang

2009-01-01

The viscosity is rate-limiting for many processes in glassy materials such as homogenisation and crystallisation. Changes in the viscous flow behaviour in conditions of long-term irradiation are of particular interest for glassy materials used in nuclear installations as well as for nuclear waste immobilising glasses. We analyse the viscous flow behaviour of oxide amorphous materials in conditions of electron-irradiation using the congruent bond lattice model of oxide materials accounting for the flow-mediating role of broken bonds termed configurons. An explicit equation of viscosity was obtained which is in agreement with experimental data for non-irradiated glasses and shows for irradiated glasses, first, a significant decrease of viscosity, and, second, a stepwise reduction of the activation energy of flow. An equation for glass-transition temperature was derived which shows that irradiated glasses have lower glass transition temperatures. Intensive electron irradiation of glasses causes their fluidization due to non-thermal bond breaking and can occur below the glass transition temperature. Due to surface tension forces fluidization of glasses at enough high electron flux densities can result in modification of nano-size volumes and particles such as those experimentally observed under TEM electron beams. (authors)

Waste glass melting stages

International Nuclear Information System (INIS)

Anderson, L.D.; Dennis, T.; Elliott, M.L.; Hrma, P.

1993-04-01

Three different simulated nuclear waste glass feeds, consisting of dried waste and glass frit, were heat treated for 1 hour in a gradient furnace at temperatures ranging from approximately 600 degrees C--1000 degrees C. Simulated melter feeds from the Hanford Waste Vitrification Plant (HWVP), the Defense Waste Processing Facility (DWPF), and Kernforschungszentrum Karlsruhe (KfK) in Germany were used. The samples were thin-sectioned and examined by optical microscopy to investigate the stages of the conversion from feed to glass. Various phenomena were seen, such as frit softening, bubble formation, foaming, bubble motion and removal, convective mixing, and homogenization. Behavior of different feeds was similar, although the degree of gas generation and melt homogenization varied

Microporous glasses for pro-ecological applications

International Nuclear Information System (INIS)

Procyk, B.; Stoch, L.; Kubacki, M.; Rewilak, M.; Soltysiak, J.

1994-01-01

Microporous glasses are obtained by appropriate thermal and chemical treatment. On account of their sorptive properties microporous glasses represent an excellent material for storing high-energy radioactive waste products in nuclear power engineering and for binding toxins in natural environment. Microporous glasses may be used as membranes non-organic, as gel filling in chromatography. They may find application in biochemistry, chemical, metallurgical, electrochemical and other industries. The above applications depend on the internal arrangement, size and shape of pores. (author). 4 refs, 4 figs, 1 tab

Conversion of Nuclear Waste into Nuclear Waste Glass: Experimental Investigation and Mathematical Modeling

International Nuclear Information System (INIS)

Hrma, Pavel

2014-01-01

The melter feed, slurry, or calcine charged on the top of a pool of molten glass forms a floating layer of reacting material called the cold cap. Between the cold-cap top, which is covered with boiling slurry, and its bottom, where bubbles separate it from molten glass, the temperature changes by up to 1000 K. The processes that occur over this temperature interval within the cold cap include liberation of gases, conduction and consumption of heat, dissolution of quartz particles, formation and dissolution of intermediate crystalline phases, and generation of foam and gas cavities. These processes have been investigated using thermal analyses, optical and electronic microscopies, x-ray diffraction, as well as other techniques. Properties of the reacting feed, such as heat conductivity and density, were measured as functions of temperature. Investigating the structure of quenched cold caps produced in a laboratory-scale melter complemented the crucible studies. The cold cap consists of two main layers. The top layer contains solid particles dissolving in the glass-forming melt and open pores through which gases are escaping. The bottom layer contains bubbly melt or foam where bubbles coalesce into larger cavities that move sideways and release the gas to the atmosphere. The feed-to-glass conversion became sufficiently understood for representing the cold-cap processes via mathematical models. These models, which comprise heat transfer, mass transfer, and reaction kinetics models, have been developed with the final goal to relate feed parameters to the rate of glass melting

Effect of geologic repository parameters on aqueous corrosion of nuclear glass

International Nuclear Information System (INIS)

Tovena, I.; Advocat, T.; Jollivet, P.; Godon, N.; Vernaz, E.

1995-01-01

Twenty alumino-borosilicate glass compositions containing simulated fission product oxides were defined using the experimentation plan methodology. Three additional glass compositions were also tested. Monolithic glass corrosion tests in a dilute aqueous medium at 90 deg C indicated the variation range for the initial corrosion rates. Significant but only qualitative correlations were established between the initial corrosion rate and the molar fraction of glass network forming oxides (SiO 2 + Al 2 O 3 ), and between the initial rate and the (Na 2 O + Li 2 O + B 2 O 3 ) / (SiO 2 + Al 2 O 3 ) molar ratio in the glass. The experimentation plan allowed a polynomial model to be defined relating the initial corrosion rate at 90 deg C to the oxide concentrations in the glass. Although the model is theoretically capable of predicting the corrosion rates, it does not always account for the actual data measured during other experiments; this discrepancy may be attributable either to the presence of other chemical elements (MgO) or to CaO concentrations differing from the fixed value adopted for the experimentation plan. Glass powder corrosion tests designed to simulate advanced corrosion reaction progress, account for the wide variations in the dissolved glass quantities, although no correlation exists with the glass chemical composition. (authors). 49 refs., 4 figs., 34 tabs

Effect of geologic repository parameters on aqueous corrosion of nuclear glass

Energy Technology Data Exchange (ETDEWEB)

Tovena, I; Advocat, T; Jollivet, P; Godon, N; Vernaz, E

1996-12-31

Twenty alumino-borosilicate glass compositions containing simulated fission product oxides were defined using the experimentation plan methodology. Three additional glass compositions were also tested. Monolithic glass corrosion tests in a dilute aqueous medium at 90 deg C indicated the variation range for the initial corrosion rates. Significant but only qualitative correlations were established between the initial corrosion rate and the molar fraction of glass network forming oxides (SiO{sub 2} + Al{sub 2}O{sub 3}), and between the initial rate and the (Na{sub 2}O + Li{sub 2}O + B{sub 2}O{sub 3}) / (SiO{sub 2} + Al{sub 2}O{sub 3}) molar ratio in the glass. The experimentation plan allowed a polynomial model to be defined relating the initial corrosion rate at 90 deg C to the oxide concentrations in the glass. Although the model is theoretically capable of predicting the corrosion rates, it does not always account for the actual data measured during other experiments; this discrepancy may be attributable either to the presence of other chemical elements (MgO) or to CaO concentrations differing from the fixed value adopted for the experimentation plan. Glass powder corrosion tests designed to simulate advanced corrosion reaction progress, account for the wide variations in the dissolved glass quantities, although no correlation exists with the glass chemical composition. (authors). 49 refs., 4 figs., 34 tabs.

Application of the NNWSI [Nevada Nuclear Waste Storage Investigations] unsaturated test method to actinide doped SRL [Savannah River Laboratory] 165 type glass

International Nuclear Information System (INIS)

Bates, J.K.; Gerding, T.J.

1990-08-01

The results of tests done using the Unsaturated Test Method are presented. These tests, done to determine the suitability of glass in a potential high-level waste repository as developed by the Nevada Nuclear Waste Storage Investigations Project, simulate conditions anticipated for the post-containment phase of the repository when only limited contact between the waste form and water is expected. The reaction of glass occurs via processes that are initiated due to glass/water vapor and glass/liquid water contact. Vapor interaction results in the initiation of an exchange process between water and the more mobile species (alkalis and boron) in the glass. The liquid reaction produces interactions similar to those seen in standard leaching tests, except due to the limited amount of water present and the presence of partially sensitized 304L stainless steel, the formation of reaction products greatly exceeds that found in MCC-1 type leach tests. The effect of sensitized stainless steel on the reaction is to enhance breakdown of the glass matrix thereby increasing the release of the transuranic elements from the glass. However, most of the Pu and Am released is entrained by either the metal components of the test or by the reaction phases, and is not released to solution. 16 refs., 20 figs., 17 tabs

Fracture during cooling of cast borosilicate glass containing nuclear wastes

International Nuclear Information System (INIS)

Smith, P.K.; Baxter, C.A.

1981-09-01

Procedures and techniques were evaluated to mitigate thermal stress fracture in waste glass as the glass cools after casting. The two principal causes of fracture identified in small-scale testing are internal thermal stresses arising from excessive thermal gradients when cooled too fast, and shear fracturing in the surface of the glass because the stainless steel canister shrinks faster than the glass on cooling. Acoustic emission and ceramographic techniques were used to outline an annealing schedule that requires at least three weeks of controlled cooling below 550 0 C to avoid excessive thermal gradients and corresponding stresses. Fracture arising from canister interactions cannot be relieved by slow cooling, but can be eliminated for stainless steel canisters by using ceramic paper, ceramic or graphite paste linings, or by choosing a canister material with a thermal expansion coefficient comparable to, or less than, that of the glass

Borosilicate glass for gamma irradiation fields

Science.gov (United States)

Baydogan, N.; Tugrul, A. B.

2012-11-01

Four different types of silicate glass specimens were irradiated with gamma radiation using a Co-60 radioisotope. Glass specimens, with four different chemical compositions, were exposed to neutron and mixed neutron/gamma doses in the central thimble and tangential beam tube of the nuclear research reactor. Optical variations were determined in accordance with standardisation concept. Changes in the direct solar absorbance (αe) of borosilicate glass were examined using the increase in gamma absorbed dose, and results were compared with the changes in the direct solar absorbance of the three different type silicate glass specimens. Solar absorption decreased due to decrease of penetration with absorbed dose. αe of borosilicate increased considerably when compared with other glass types. Changes in optical density were evaluated as an approach to create dose estimation. Mixed/thermal neutron irradiation on glass caused to increse αe.

Application of the GRAAL model to leaching experiments with SON68 nuclear glass in initially pure water

International Nuclear Information System (INIS)

Frugier, P.; Chave, T.; Gin, S.; Lartigue, J.-E.

2009-01-01

Based on a review of the current state of knowledge concerning the aqueous alteration of SON68 nuclear glass we have proposed a mechanistic model, GRAAL (Glass Reactivity with Allowance for the Alteration Layer) [P. Frugier, S. Gin, Y. Minet, T. Chave, B. Bonin, N. Godon, J.E. Lartigue, P. Jollivet, A. Ayral, L. De Windt, G. Santarini, J. Nucl. Mater. 380 (2008) 8]. This article describes how the GRAAL model hypotheses are solved using a calculation code coupling chemistry and transport. The geochemical solution of this model combines three major phenomena: chemical equilibria in solution, water and ion transport by convection or diffusion, and element diffusion through the passivating reactive interphase. The model results are compared with experimental data for SON68 glass leached in initially pure water both in a closed system and in renewed media. The comparison shows the model very satisfactorily accounts for variations in the pH and the element concentrations in solution as a function of time, the glass surface area in contact with solution, and the solution renewal rate. This success is due to the fact that the diffusion of elements through the alteration gel is taken into account in the model. This mechanism cannot be disregarded under most experimental conditions - if only to predict the solution pH - and must therefore be an integral part of the geochemical model.

Dissolution of basaltic glass in seawater: Mechanism and rate

International Nuclear Information System (INIS)

Crovisier, J.L.; Honnorez, J.; Eberhart, J.P.

1987-01-01

Basaltic glasses are considered as natural analogues for nuclear waste glasses. Thermodynamic computer codes used to evaluate long term behavior of both nuclear waste and basaltic glasses require the knowledge of the dissolution mechanism of the glass network. The paper presents the results of a series of experiments designed to study the structure and chemical composition of alteration layers formed on the surface of artificial tholeiitic glass altered in artificial seawater. Experiments were performed at 60 degree C, 1 bar and 350 bars in non-renewed conditions. A natural sample from Palagonia (Sicily) has been studied by electron microscopy and comparison between natural and experimental palagonitic layers is made. The behavior of dissolved silica during experiments, and both the structure and the chemical composition of the palagonitic layers, indicate that they form by precipitation of secondary minerals from solution after a total breakdown of the glassy network, i.e., congruent dissolution of the glass. Hence the dissolution equation necessary for thermodynamic modelling of basaltic glass dissolution in seawater at low temperature must be written as a simple stoichiometric process. These experiments indicate that the transformation of glass to palagonitic material is not isovolumetric. Hence it is preferable to use Fe or Ti as conservative elements for chemical budget calculations

Low leach rate glasses for immobilization of nuclear wastes

International Nuclear Information System (INIS)

Chick, L.A.; Buckwalter, C.Q.

1980-10-01

Improved defense and commercial waste glass have about one order of magnitude lower leach rates at 90 0 C in static deionized water than reference glasses. This durability difference diminishes as the leaching temperature is raised, but at repository temperature less than 150 0 C, the improved compositions would have considerable advantages over reference glases. At the melting temperatures necessary for most of the high-durability glasses, volatility was found to be higher than that experienced in processing current reference glases. Higher volatilities might be compensated for by specific design of the off-gas system for improved off-gas treatment and volatile materials recovery. 6 figures, 2 tables

Nuclear microprobe analysis of carbon within glass inclusions and volcanic materials

International Nuclear Information System (INIS)

Metrich, N.; Mosbah, M.; Trocellier, P.; Clocchiatti, R.

1986-01-01

Microanalysis possibilities have been explored to determine light element concentrations within glasses (melt inclusions and basaltic glass fragments) and volcanic phenocrysts. In the first step, C was examined. The study of different spectral interferences lead to calculated detection limits of 40 μg/g for basaltic glasses and 50 μg/g for olivine crystals. The C contents of all investigated specimens range from 40 μg/g (the detection limit) to 6800 μg/g. Heterogeneities were revealed within glass inclusions. Measurements show obvious concentration profiles in basaltic glass samples. Our results agree with previous published data and are reliable. Accuracy of measurements is about 20%. 12 refs

Durability testing with West Valley borosilicate glass composition- Phase II

International Nuclear Information System (INIS)

Macedo, P.B.; Finger, S.M.; Barkatt, A.A.; Pegg, I.L.; Feng, X.; Freeborn, W.P.

1988-06-01

This report presents the research performed by the Catholic University of America Vitreous State Laboratory (VSL) during FY 1987 in support of the West Valley Demonstration Project (WVDP) nuclear waste vitrification process. A principal objective of this work is the optimization of the glass composition be used for the vitrification of the liquid high-level waste generated at West Valley during nuclear fuel reprocessing. This report discusses (1) the experimental investigations to optimize the reference glass composition (the current leading candidates are WVCM-50 and ATM-10) for the WVDP vitrification process; (2) the systematic experimental investigation performed to determine the effects of compositional variations in WVCM-50 and WV-205 reference glasses on their viscosity and durability (including initial results of long-term leach tests of WVCM-50 under repository conditions); (3) the development of short-time and predictive leach tests; (4) the development of a process model for the West Valley vitrification process which predicts the range of glass compositions which may be encountered during normal operations and the effects of deviations in process control parameters; and (5) the development of product models for predicting the durability and viscosity of nuclear waste glasses

Effects of radionuclide decay on waste glass behavior: A critical review

International Nuclear Information System (INIS)

Wronkiewicz, D.J.

1993-12-01

This paper is an extension of a chapter in an earlier report [1] that provides an updated review on the status of radiation damage problems in nuclear waste glasses. This report will focus on radiation effects on vitrified borosilicate nuclear waste glasses under conditions expected in the proposed Yucca mountain repository. Radiation effects on high-level waste glasses and their surrounding repository environment are important considerations for radionuclide immobilization because of the potential to alter the glass stability and thereby influence the radionuclide retentive properties of this waste form. The influence of radionuclide decay on vitrified nuclear waste may be manifested by several changes, including volume, stored energy, structure, microstructure, mechanical properties, and phase separation. Radiation may also affect the composition of aqueous fluids and atmospheric gases in relatively close proximity to the waste form. What is important to the radionuclide retentive properties of the repository is how these radiation effects collectively or individually influence the durability and radionuclide release from the glass in the event of liquid water contact

Radiation shielding application of lead glass

International Nuclear Information System (INIS)

Nathuram, R.

2017-01-01

Nuclear medicine and radiotherapy centers equipped with high intensity X-ray or teletherapy sources use lead glasses as viewing windows to protect personal from radiation exposure. Lead is the main component of glass which is responsible for shielding against photons. It is therefore essential to check the shielding efficiency before they are put in use. This can be done by studying photon transmission through the lead glasses. The study of photon transmission in shielding materials has been an important subject in medical physics and is potential useful in the development of radiation shielding materials

Durable Glass For Thousands Of Years

International Nuclear Information System (INIS)

Jantzen, C.

2009-01-01

The durability of natural glasses on geological time scales and ancient glasses for thousands of years is well documented. The necessity to predict the durability of high level nuclear waste (HLW) glasses on extended time scales has led to various thermodynamic and kinetic approaches. Advances in the measurement of medium range order (MRO) in glasses has led to the understanding that the molecular structure of a glass, and thus the glass composition, controls the glass durability by establishing the distribution of ion exchange sites, hydrolysis sites, and the access of water to those sites. During the early stages of glass dissolution, a 'gel' layer resembling a membrane forms through which ions exchange between the glass and the leachant. The hydrated gel layer exhibits acid/base properties which are manifested as the pH dependence of the thickness and nature of the gel layer. The gel layer ages into clay or zeolite minerals by Ostwald ripening. Zeolite mineral assemblages (higher pH and Al 3+ rich glasses) may cause the dissolution rate to increase which is undesirable for long-term performance of glass in the environment. Thermodynamic and structural approaches to the prediction of glass durability are compared versus Ostwald ripening.

DURABLE GLASS FOR THOUSANDS OF YEARS

Energy Technology Data Exchange (ETDEWEB)

Jantzen, C.

2009-12-04

The durability of natural glasses on geological time scales and ancient glasses for thousands of years is well documented. The necessity to predict the durability of high level nuclear waste (HLW) glasses on extended time scales has led to various thermodynamic and kinetic approaches. Advances in the measurement of medium range order (MRO) in glasses has led to the understanding that the molecular structure of a glass, and thus the glass composition, controls the glass durability by establishing the distribution of ion exchange sites, hydrolysis sites, and the access of water to those sites. During the early stages of glass dissolution, a 'gel' layer resembling a membrane forms through which ions exchange between the glass and the leachant. The hydrated gel layer exhibits acid/base properties which are manifested as the pH dependence of the thickness and nature of the gel layer. The gel layer ages into clay or zeolite minerals by Ostwald ripening. Zeolite mineral assemblages (higher pH and Al{sup 3+} rich glasses) may cause the dissolution rate to increase which is undesirable for long-term performance of glass in the environment. Thermodynamic and structural approaches to the prediction of glass durability are compared versus Ostwald ripening.

Challenges in commercial manufacture of radiation shielding glasses

International Nuclear Information System (INIS)

Gupta, R.K.

2011-01-01

Radioactive hot-cells employ Radiation Shielding Windows (RSWs), assembled from specialty glasses, developed exclusively for nuclear industry. RSWs serve the twin purpose of direct viewing and shielding protection to the operator and use various types of radiation resistant and optically compatible glasses, such as low-density borosilicate glass; medium-density glass with up to 45% Lead and high-density glass with over 70% lead. Some glasses are Ceria-doped for enhancing their resistance threshold to radiation browning. A clear view of future requirement, capital and environmental costs could be the driving force towards bringing about changes in melting practices, encourage melting development, and enhancing collaboration. With DAE and CGCRI working in tandem, production of the entire range of RSW glasses by an Indian glass industry participant may no longer be a distant dream

Structural changes in irreversibly densified fused silica: implications for the chemical resistance of high level nuclear waste glasses

International Nuclear Information System (INIS)

Susman, S.; Volin, K.J.; Liebermann, R.C.; Gwanmesia, G.D.; Yanbin Wang

1990-01-01

Energetic photons and energetic particles create changes in the structure of nuclear waste glasses. These can be observed as changes in the average bulk physical properties. For example, exposure of fused silica to high doses of neutron bombardment leads to a maximum average compaction of 3%. However, this does not reveal the true extent of the densification that takes place at a microscopic level. Recent advances in high pressure technology have yielded large samples of fused silica which have been permanently densified under pressure and whose bulk density has been increased by 20%. These specimens have an overall structure that replicates the microstructure of a radiation damaged glass. Measurements have been made for the first time of the structural changes in this pressure densified vitreous silica using neutron diffraction and infrared absorption spectrometry. Extensive alterations in intermediate range order have been observed with consequent anticipated changes in chemical reactivity. The resistance of high level waste glasses to leaching by groundwater must be considered in light of these experimental findings. (author)

Alteration of nuclear glass in contact with iron and claystone at 90 °C under anoxic conditions: Characterization of the alteration products after two years of interaction

International Nuclear Information System (INIS)

Schlegel, Michel L.; Martin, Christelle; Brucker, Florence; Bataillon, Christian; Blanc, Cécile; Chorro, Matthieu; Jollivet, Patrick

2016-01-01

The present study investigates the alteration of a fractured glass block in contact with iron and Callovo-Oxfordian claystone at 90 °C under anoxic and water-saturated conditions. The alteration rates and the nature of glass alteration products at the different compact interfaces (glass-clay, glass-iron) and in cracks were assessed by solution chemistry and microscopic-scale techniques (scanning electron microscopy coupled with energy-dispersive X-ray microscopy, microRaman spectroscopy, and X-ray absorption fine structure spectroscopy). A significant but modest (two-fold) increase in glass alteration in contact with steel was observed, leading to an average alteration rate over the experiment of about 0.007–0.014 g/m"2/d. This rate is significantly lower than forward rate r_0 in clay-equilibrated groundwater (1.7 g/m"2/d), indicating that a decrease of the alteration rate was not hindered by the steel presence. The corrosion–alteration interface was made up of successive layers of corrosion products in contact with iron, a layer of Fe silicates, and an altered glass layer enriched in Fe. Characterization of the glass block in direct contact with claystone revealed that the thickness of altered glass was much more important than at the glass-iron interface. The altered glass layer in contact with clay was slightly enriched in Fe and Mg, and depleted in alkali cations. Altered glass layers in cracks were usually limited to fringes thinner than 2 μm, with a thickness decreasing from the crack mouth, indicating that alteration is controlled by transport in the cracks. The fractures were partially filled with calcite and lanthanide hydroxocarbonate precipitates. These results contribute to the understanding of nuclear vitrified waste-iron-corrosion products interactions in a deep geological repository. - Highlights: • Anoxic alteration of glass in contact with iron and clay at 90 °C for two-years. • Alteration rates of 0.015 and 0.5 g/m"2/d at glass

Nuclear waste disposal: alternatives to solidification in glass proposed

International Nuclear Information System (INIS)

Kerr, R.A.

1979-01-01

More than a quarter-million cubic meters of liquid radioactive wastes are now being held at government installations awaiting final disposal. During the past 20 years, the disposal plan of choice has been to incorporate the 40 to 50 radioactive elements dissolved in liquid wastes into blocks of glass, seal the glass in metal canisters, and insert the canisters into deep, geologically stable salt beds. Over the last few years, some geologists and materials scientists have become concerned that perhaps not enough is known yet about the interaction of waste, container, and salt (or any rock) to have a reasonable assurance that the hazardous wastes will be contained successfully. The biggest advantage of glass at present is the demonstrated practicality of producing large, highly radioactive blocks of it. The frontrunner as a successor to glass is ceramics, which are nonmetallic crystalline materials formed at high temperature, such as chinaware or natural minerals. An apparent advantage of ceramics is that they already have an ordered atomic structure, whose properties can be tailored to a particular waste element and to conditions of a specific disposal site. A ceramic tailored for waste disposal called supercalcine-ceramic has been developed. It was emphasized that the best minerals for waste solidification may be those that have proved most stable under natural conditions over geologic time. Disadvantage to ceramics are radiation damage and transmutation. However, it is now obvious that some ceramics are more stable than glass under certain conditions. Metal-encapsulated ceramic, called cermet, is being developed as a waste form. Cermets are considerably more resistant at 100 0 C than a borosilicate waste glass. Researchers are now testing prospective waste forms under the most extreme conditions that might prevail in a waste disposal site

High-level waste glass compendium; what it tells us concerning the durability of borosilicate waste glass

International Nuclear Information System (INIS)

Cunnane, J.C.; Allison, J.

1993-01-01

Facilities for vitrification of high-level nuclear waste in the United States are scheduled for startup in the next few years. It is, therefore, appropriate to examine the current scientific basis for understanding the corrosion of high-level waste borosilicate glass for the range of service conditions to which the glass products from these facilities may be exposed. To this end, a document has been prepared which compiles worldwide information on borosilicate waste glass corrosion. Based on the content of this document, the acceptability of canistered waste glass for geological disposal is addressed. Waste glass corrosion in a geologic repository may be due to groundwater and/or water vapor contact. The important processes that determine the glass corrosion kinetics under these conditions are discussed based on experimental evidence from laboratory testing. Testing data together with understanding of the long-term corrosion kinetics are used to estimate radionuclide release rates. These rates are discussed in terms of regulatory performance standards

Comparison of the corrosion behaviors of the glass-bonded sodalite ceramic waste form and reference HLW glasses

International Nuclear Information System (INIS)

Ebert, W. L.; Lewis, M. A.

1999-01-01

A glass-bonded sodalite ceramic waste form is being developed for the long-term immobilization of salt wastes that are generated during spent nuclear fuel conditioning activities. A durable waste form is prepared by hot isostatic pressing (HIP) a mixture of salt-loaded zeolite powders and glass frit. A mechanistic description of the corrosion processes is being developed to support qualification of the CWF for disposal. The initial set of characterization tests included two standard tests that have been used extensively to study the corrosion behavior of high level waste (HLW) glasses: the Material Characterization Center-1 (MCC-1) Test and the Product Consistency Test (PCT). Direct comparison of the results of tests with the reference CWF and HLW glasses indicate that the corrosion behaviors of the CWF and HLW glasses are very similar

Waste glass melting stages

International Nuclear Information System (INIS)

Anderson, L.D.; Dennis, T.; Elliott, M.L.; Hrma, P.

1994-01-01

Three simulated nuclear waste glass feeds, consisting of dried waste and glass frit, were heat treated for 1 hour in a gradient furnace at temperatures ranging from approximately 600 degrees C to 1000 degrees C. Simulated melter feeds from the Hanford Waste Vitrification Plant (HWVP), the Defense Waste Processing Facility (DWPF), and Kernforschungszentru Karlsruhe (KfK) in Germany were used. The samples were thin sectioned and examined by optical microscopy to investigate the stages of the conversion from feed to glass. Various phenomena were seen, such as frit softening, bubble formation, foaming, bubble motion and removal, convective mixing, and homogenization. The behavior of different feeds was similar, although the degree of gas generation and melt homogenization varied. 2 refs., 8 tabs

Porous glass matrix method for encapsulating high-level nuclear wastes

International Nuclear Information System (INIS)

Macedo, P.B.; Tran, D.C.; Simmons, J.H.; Saleh, M.; Barkatt, A.; Simmons, C.J.; Lagakos, N.; DeWitt, E.

1979-01-01

A novel process which uses solidified porous high-silica glass powder to fixate radioactive high-level wastes is described. The process yields cylinders consisting of a core of high-silica glass containing the waste elements in its structure and a protective layer also of high-silica glass completely free of waste elements. The process can be applied to waste streams containing 0 to 100% solids. The core region exhibits a higher coefficient of thermal expansion and a lower glass transition temperature than the outer protective layer. This leads to mechanical strengthening of the glass and good resistance to stress corrosion by the development of a high residual compressive stress on the surface of the sample. Both the core and the protective layer exhibit extremely high chemical durability and offer an effective fixation of the radioactive waste elements, including 239 Pu and 99 Tc which have long half-lives, for calculated periods of more than 1 million years, when temperatures are not allowed to rise above 100 0 C

R7T7 nuclear waste glass behavior in moist clay: role of the clay mass/glass surface area ratio

International Nuclear Information System (INIS)

Godon, N.; Vernaz, E.

1989-01-01

R7T7 glass alteration was investigated in the presence of various moist clays. In contact with smectite 4a, selected in France as a potential engineered barrier material, the glass was significantly corroded: after 6 months the glass corrosion rate was practically the same as the initial alteration rate in double-distilled water. Substantially lower alteration was observed in contact with bentonite 6 activated by sodium carbonate. Smectite 4a consumes silicon released by glass corrosion, and thus retards the rise to high silicon concentrations in solution at which the glass corrosion rate diminishes. Glass can therefore in some cases be altered as much in moist clay as in water with high renewal rates. Other experiments with smaller quantities of smectite 4a showed that the phenomenon becomes less important in time: its duration is proportional to the ratio between the clay mass (C) and the glass surface area (SA). Comparing the results of studies at different C/SA ratios indicates that low glass corrosion rates are obtained more slowly at higher C/SA ratios. Tests with 239 Pu-doped R7T7 glass also showed that the radionuclide retention factor in the alteration film at the glass surface is only 6, compared with a factor of nearly 50 in double-distilled water

Phase formation during corrosion experiments with two simulated borosilicate nuclear waste glasses

International Nuclear Information System (INIS)

Haaker, R.F.

1985-10-01

Corrosion products resulting from the reaction of simulated high-level radioactive waste glasses with various solutions have been identified. At 200degC, in saturated NaCl, a degree of reaction of 10 g C31-3 glass or 2.6 g SON 68 glass per liter of solution was obtained. Analcime, vermiculite (a phyllosilicate) and a 2:1 zinc silicate are the major silica containing alteration products for the C31-3 glass. Analcime was the only silicate alteration product which could be identified for SON 68 glass. C31-3 glass appeared to be less reactive with a quinary brine containing Mg ++ than with NaCl. With the quinary brine, montmorillonite (a phyllosilicate) was the predominant silica containing alteration product. Hydrotalcite (a Mg-Al hydroxysulfate) and montmorillonite were the major Al-containing phases. A phyllosilicate, probably montmorillonite, was observed to form during the reaction of SON 68 glass with quinary brine. With either glass, modified NaCl brines which contained small amounts of MgCl 2 seem to have the effect of decreasing the amount of analcime and increasing the amount of phyllosilicate which is formed. In the case of C31-3 glass, there is approximately enough Mg, Al and Zn to precipitate most of the leached Si; measured Si concentrations remain well below that expected for amorphous silica. SON 68 glass has less Zn, Al and Mg than C31-3 glass and much higher Si concentrations of the leachates. (orig./RB)

Fluorescent nuclear track images of Ag-activated phosphate glass irradiated with photons and heavy charged particles

Energy Technology Data Exchange (ETDEWEB)

Kurobori, Toshio, E-mail: kurobori@staff.kanazawa-u.ac.jp [Graduate School of Natural Science and Technology, Kanazawa University, Kakuma, Kanazawa 920-1192 (Japan); Yanagida, Yuka [Oarai Research Center, Chiyoda Technol Corporation, Oarai-machi, Ibaraki 311-1313 (Japan); Kodaira, Satoshi [National Institute of Radiological Sciences, 4-9-1 Anagawa, Inage-ku, Chiba 263-8555 (Japan); Shirao, Taichi [Nikon Instech Co., Ltd., Tanakanishi, Sakyo-ku, Kyoto 606-8221 (Japan)

2017-05-21

In this paper we report about the demonstration of the nuclear track imaging capabilities of Ag-activated phosphate glass. A 375 nm laser and confocal laser scanning microscopy (CLSM) were respectively used for track excitation and detection. Specifically, the blue and orange radiophotoluminescent (RPL) tracks and dose distributions observed after irradiation with soft X-rays, gamma rays and heavy charged particles (HCPs) are examined. In addition, the origins of the reductions in RPL efficiency for high-dose X-ray irradiation and for irradiation with HCPs with high linear energy transfer (LET) values are investigated via a CLSM and a conventional fluorescent reader and discussed. - Highlights: • 3D track images are demonstrated using a confocal laser microscopy. • Fluorescent track detectors are based on RPL Ag-doped phosphate glass. • The dose distributions are examined for X-ray, gamma ray and HCP irradiations. • The origins of the reduction in RPL efficiency are investigated and discussed.

Fluorescent nuclear track images of Ag-activated phosphate glass irradiated with photons and heavy charged particles

International Nuclear Information System (INIS)

Kurobori, Toshio; Yanagida, Yuka; Kodaira, Satoshi; Shirao, Taichi

2017-01-01

In this paper we report about the demonstration of the nuclear track imaging capabilities of Ag-activated phosphate glass. A 375 nm laser and confocal laser scanning microscopy (CLSM) were respectively used for track excitation and detection. Specifically, the blue and orange radiophotoluminescent (RPL) tracks and dose distributions observed after irradiation with soft X-rays, gamma rays and heavy charged particles (HCPs) are examined. In addition, the origins of the reductions in RPL efficiency for high-dose X-ray irradiation and for irradiation with HCPs with high linear energy transfer (LET) values are investigated via a CLSM and a conventional fluorescent reader and discussed. - Highlights: • 3D track images are demonstrated using a confocal laser microscopy. • Fluorescent track detectors are based on RPL Ag-doped phosphate glass. • The dose distributions are examined for X-ray, gamma ray and HCP irradiations. • The origins of the reduction in RPL efficiency are investigated and discussed.

First principles process-product models for vitrification of nuclear waste: Relationship of glass composition to glass viscosity, resistivity, liquidus temperature, and durability

International Nuclear Information System (INIS)

Jantzen, C.M.

1991-01-01

Borosilicate glasses will be used in the USA and in Europe to immobilize radioactive high level liquid wastes (HLLW) for ultimate geologic disposal. Process and product quality models based on glass composition simplify the fabrication of the borosilicate glass while ensuring glass processability and quality. The process model for glass viscosity is based on a relationship between the glass composition and its structural polymerization. The relationship between glass viscosity and electrical resistivity is also shown to relate to glass polymerization. The process model for glass liquidus temperature calculates the solubility of the liquidus phases based on the free energies of formation of the precipitating species. The durability product quality model is based on the calculation of the thermodynamic hydration free energy from the glass composition

Contributions of vitreous natural analogs to the investigation of long-term nuclear glass behavior; Apports des analogues naturels vitreux a la validation des codes de prediction du comportement a long terme des verres nucleaires

Energy Technology Data Exchange (ETDEWEB)

Techer, I

1999-07-01

This study assesses the extend of the analogy between the alteration behavior in water and in a moist clay environment of aluminosilicate volcanic glass and alumino-borosilicate nuclear containment glass. Basaltic glass alteration in water initially occurs by hydrolysis processes with an activation energy on the order of 73 kJ.mol{sup -1}. As the reaction progresses, the alteration rate drops by over four orders of magnitude from the initial rate r{sub 0}, The alteration kinetics are not governed by the alteration solution chemistry alone, the glass alteration film appears to have a major role as a diffusion barrier limiting the transfer of reaction species and products. All these aspects highlight the behavioral analogy between basaltic glass and nuclear borosilicate glass in aqueous media. Conversely, the alteration reaction of obsidian-type volcanic glass involves other mechanisms than those governing the dissolution of borosilicate glass. Basaltic glass alteration is also examined in the presence of a clay environmental material, in a study of the natural basaltic glass and argillaceous pelites system of the Salagou basin in southern France, in an approach combining mineralogical, chemical and isotopic data to assess the interactions between a basaltic glass and the argillaceous pelites. Laboratory leach test results with basaltic glass and measured data for the Salagou glass in its natural environment are modeled using a code implementing a kinetic law coupling diffusive transfer of dissolved silica with a reaction affinity law. (author)

The corrosion behavior of DWPF glasses

International Nuclear Information System (INIS)

Ebert, W.L.; Bates, J.K.

1995-01-01

The authors analyzed the corroded surfaces of reference glasses developed for the Defense Waste Processing Facility (DWPF) to characterize their corrosion behavior. The corrosion mechanism of nuclear waste glasses must be known in order to provide source terms describing radionuclide release for performance assessment calculations. Different DWPF reference glasses were corroded under conditions that highlighted various aspects of the corrosion process and led to different extents of corrosion. The glasses corroded by similar mechanisms, and a phenomenological description of their corrosion behavior is presented here. The initial leaching of soluble glass components results in the formation of an amorphous gel layer on the glass surface. The gel layer is a transient phase that transforms into a layer of clay crystallites, which equilibrates with the solution as corrosion continues. The clay layer does not act as a barrier to either water penetration or glass dissolution, which continues beneath it, and may eventually separate from the glass. Solubility limits for glass components may be established by the eventual precipitation of secondary phases; thus, corrosion of the glass becomes controlled by the chemical equilibrium between the solution and the assemblage of secondary phases. In effect, the solution is an intermediate phase through which the glass transforms to an energetically more favorable assemblage of phases. Implications regarding the prediction of long-term glass corrosion behavior are discussed

Effect of alumina on the dissolution rate of glasses

International Nuclear Information System (INIS)

Palavit, G.; Montagne, L.

1997-01-01

Small alumina addition to silicate glasses improves their chemical durability, but a large amount of alumina can also be beneficial to obtain a high dissolution rate. This paper describes the effect of Al 3+ on the early stage of glass alteration, in relation with its coordination in the glass and also with the reactions involved (hydrolysis and ionic exchange). We describe briefly nuclear magnetic resonance tools available to characterize the aluminum environments in the glasses. The rote of alumina on the dissolution rate of phosphate glasses is also discussed in order to show that the effect of Al 3+ is dependant upon the nature of the glass matrix. (author)

Vitrification chemistry and nuclear waste

International Nuclear Information System (INIS)

Plodinec, M.J.

1985-01-01

The vitrification of nuclear waste offers unique challenges to the glass technologist. The waste contains 50 or 60 elements, and often varies widely in composition. Most of these elements are seldom encountered in processing commercial glasses. The melter to vitrify the waste must be able to tolerate these variations in composition, while producing a durable glass. This glass must be produced without releasing hazardous radionuclides to the environment during any step of the vitrification process. Construction of a facility to convert the nearly 30 million gallons of high-level nuclear waste at the Savannah River Plant into borosilicate glass began in late 1983. In developing the vitrification process, the Savannah River Laboratory has had to overcome all of these challenges to the glass technologist. Advances in understanding in three areas have been crucial to our success: oxidation-reduction phenomena during glass melting; the reaction between glass and natural wastes; and the causes of foaming during glass melting

Alteration of rhyolitic (volcanic) glasses in natural Bolivian salt lakes. - Natural analogue for the behavior of radioactive waste glasses in rock salt repositories

International Nuclear Information System (INIS)

Abdelouas, A.

1996-06-01

Alteration experiments with the R7T7 glass in three salt brines, saturated respectively in MgCl 2 , MgCl 2 -CaCl 2 and NaCl, showed that the solubilities of most radionuclides are controlled by the secondary phases. Nd, La, and Pr are trapped in powellite, Ce in cerianite, U in coffinite, and Sr is partially immobilized in barite. There is a good similarity between the secondary phases formed experimentally on volcanic glasses and the R7T7 glass altered in MgCl 2 CaCl 2 -saturated brine (formation of hydrotalcite and chlorite-serpentine at short-term and saponite at long-term). These results support the use of volcanic glasses alteration patterns in Mg-rich solutions (seawater, brines) to understand the long-term behavior of nuclear waste glasses and to evaluate the stability of the secondary phases. The study of the sediments of Uyuni (Bolivia) showed that the corrosion rate of the rhyolitic glass in brines at 10 C is 12 to 30 time lower than those of rhyolitic glasses altered in high dilute conditions. The neoformed phases in the sediments are: Smectite, alunite, pyrite, barite, celestite and cerianite. The low alteration rate of rhyolitic glasses in brines and the formation of secondary phases such as smectite, barite and cerianite (also formed during the experimental alteration of the R7T7 glass), permit us to expect the low alteration of nuclear waste glasses at long-term in brines and the trapping of certain radionuclides in secondary phases. (orig.) [de

Corrosion of inconel in high-temperature borosilicate glass melts containing simulant nuclear waste

Science.gov (United States)

Mao, Xianhe; Yuan, Xiaoning; Brigden, Clive T.; Tao, Jun; Hyatt, Neil C.; Miekina, Michal

2017-10-01

The corrosion behaviors of Inconel 601 in the borosilicate glass (MW glass) containing 25 wt.% of simulant Magnox waste, and in ZnO, Mn2O3 and Fe2O3 modified Mg/Ca borosilicate glasses (MZMF and CZMF glasses) containing 15 wt.% of simulant POCO waste, were evaluated by dimensional changes, the formation of internal defects and changes in alloy composition near corrosion surfaces. In all three kinds of glass melts, Cr at the inconel surface forms a protective Cr2O3 scale between the metal surface and the glass, and alumina precipitates penetrate from the metal surface or formed in-situ. The corrosion depths of inconel 601 in MW waste glass melt are greater than those in the other two glass melts. In MW glass, the Cr2O3 layer between inconel and glass is fragmented because of the reaction between MgO and Cr2O3, which forms the crystal phase MgCr2O4. In MZMF and CZMF waste glasses the layers are continuous and a thin (Zn, Fe, Ni, B)-containing layer forms on the surface of the chromium oxide layer and prevents Cr2O3 from reacting with MgO or other constituents. MgCr2O4 was observed in the XRD analysis of the bulk MW waste glass after the corrosion test, and ZrSiO4 in the MZMF waste glass, and ZrSiO4 and CaMoO4 in the CZMF waste glass.

Irradiation test of borosilicate glass burnable poison

International Nuclear Information System (INIS)

Feng Mingquan; Liao Zumin; Yang Mingjin; Lu Changlong; Huang Deyang; Zeng Wangchun; Zhao Xihou

1991-08-01

The irradiation test and post-irradiation examinations for borosilicate glass burnable poison are introduced. Examinations include visual examination, measurement of dimensions and density, and determination of He gas releasing and 10 B burnup. The corrosion and phenomenon of irradiation densification are also discussed. Two type glass samples have been irradiated with different levels of neutron flux. It proved that the GG-17 borosilicate glass can be used as burnable poison to replace the 10 B stainless steel in the Qinshan Nuclear Power Plant, and it is safe, economical and reasonable

Characterization of Analytical Reference Glass-1 (ARG-1)

International Nuclear Information System (INIS)

Smith, G.L.

1993-12-01

High-level radioactive waste may be immobilized in borosilicate glass at the West Valley Demonstration Project, West Valley, New York, the Defense Waste Processing Facility (DWPF), Aiken, South Carolina, and the Hanford Waste Vitrification Project (HWVP), Richland, Washington. The vitrified waste form will be stored in stainless steel canisters before its eventual transfer to a geologic repository for long-term disposal. Waste Acceptance Product Specifications (WAPS) (DOE 1993), Section 1.1.2 requires that the waste form producers must report the measured chemical composition of the vitrified waste in their production records before disposal. Chemical analysis of glass waste forms is receiving increased attention due to qualification requirements of vitrified waste forms. The Pacific Northwest Laboratory (PNL) has been supporting the glass producers' analytical laboratories by a continuing program of multilaboratory analytical testing using interlaboratory ''round robin'' methods. At the PNL Materials Characterization Center Analytical Round Robin 4 workshop ''Analysis of Nuclear Waste Glass and Related Materials,'' January 16--17, 1990, Pleasanton, California, the meeting attendees decided that simulated nuclear waste analytical reference glasses were needed for use as analytical standards. Use of common standard analytical reference materials would allow the glass producers' analytical laboratories to calibrate procedures and instrumentation, to control laboratory performance and conduct self-appraisals, and to help qualify their various waste forms

Simulation of alpha decay of actinides in iron phosphate glasses by ion irradiation

Energy Technology Data Exchange (ETDEWEB)

Dube, Charu L., E-mail: dubecharu@gmail.com; Stennett, Martin C.; Gandy, Amy S.; Hyatt, Neil C.

2016-03-15

Highlights: • Alpha decay of actinides in iron phosphate glasses is simulated by employing ion irradiation technique. • FTIR and Raman spectroscopic measurements confirm modification of glass network. • The depolymerisation of glass network after irradiation is attributed to synergetic effect of nuclear and electronic losses. - Abstract: A surrogate approach of ion beam irradiation is employed to simulate alpha decay of actinides in iron phosphate nuclear waste glasses. Bismuth and helium ions of different energies have been selected for simulating glass matrix modification owing to radiolysis and ballistic damage due to recoil atoms. Structural modification and change in coordination number of network former were probed by employing Reflectance Fourier-Transform Infrared (FT-IR), and Raman spectroscopies as a consequence of ion irradiation. Depolymerisation is observed in glass sample irradiated at intermediate energy of 2 MeV. Helium blisters of micron size are seen in glass sample irradiated at low helium ion energy of 30 keV.

Geochemical modelling of the long-term dissolution behaviour of the French nuclear glass R7T7

International Nuclear Information System (INIS)

Michaux, L.; Mouche, E.; Petit, J.-C.; Fritz, B.

1992-01-01

The long-term dissolution behaviour of the French nuclear reference glass R7T7 was studied by means of the geochemical code DISSOL. New experimental data which support some of the assumptions of DISSOL are presented: namely, that the dissolution is congruent and that the altered layer can be considered as an assemblage of secondary phases. At 100 o C the main results of modelling are that the altered layer is essentially formed of a pure siliceous phase (amorphous silica or chalcedony) associated with smectites and zeolites. This sequence of secondary minerals is closely linked to the chemical composition of the glass. For high degrees of reaction, corresponding to high B concentration, the ionic strength reaches 1 and the pH varies from 9 to 10 depending on the CO 2 fugacity; B,Li and Na are essentially found in solution and their concentrations depend on the amount of dissolved glass. By contrast Fe,Al and Zn have low solution concentrations which are controlled by solubility products of secondary minerals. Silicon and Ca have an intermediate behaviour which depends on the choice of selected secondary minerals. The total volume of the secondary phases is always lower than that of the corresponding dissolved glass. The results of modelling compared to static leaching experimental results show only minor differences which can be explained by kinetic control or colloid formation. It is concluded that the altered layer is not a barrier to diffusion. The consequences of this work for actinide solubility are also discussed. (author)

A mechanistic model for long-term nuclear waste glass dissolution integrating chemical affinity and interfacial diffusion barrier

Energy Technology Data Exchange (ETDEWEB)

Ma, Teqi [Northwest Institute of Nuclear Technology, No.28 Pingyu Road, Baqiao District, Xi' an,Shaanxi, 710024 (China); Mechanics and Physics of Solids Research Group, Modelling and Simulation Centre, The University of Manchester, Oxford Road, Manchester, M13 9PL (United Kingdom); Jivkov, Andrey P., E-mail: andrey.jivkov@manchester.ac.uk [Mechanics and Physics of Solids Research Group, Modelling and Simulation Centre, The University of Manchester, Oxford Road, Manchester, M13 9PL (United Kingdom); Li, Weiping; Liang, Wei; Wang, Yu; Xu, Hui [Northwest Institute of Nuclear Technology, No.28 Pingyu Road, Baqiao District, Xi' an,Shaanxi, 710024 (China); Han, Xiaoyuan, E-mail: xyhan_nint@sina.cn [Northwest Institute of Nuclear Technology, No.28 Pingyu Road, Baqiao District, Xi' an,Shaanxi, 710024 (China)

2017-04-01

Understanding the alteration of nuclear waste glass in geological repository conditions is critical element of the analysis of repository retention function. Experimental observations of glass alterations provide a general agreement on the following regimes: inter-diffusion, hydrolysis process, rate drop, residual rate and, under very particular conditions, resumption of alteration. Of these, the mechanisms controlling the rate drop and the residual rate remain a subject of dispute. This paper offers a critical review of the two most competitive models related to these regimes: affinity–limited dissolution and diffusion barrier. The limitations of these models are highlighted by comparison of their predictions with available experimental evidence. Based on the comprehensive discussion of the existing models, a new mechanistic model is proposed as a combination of the chemical affinity and diffusion barrier concepts. It is demonstrated how the model can explain experimental phenomena and data, for which the existing models are shown to be not fully adequate.

Corrosion of glass-bonded sodalite as a function of pH and temperature

International Nuclear Information System (INIS)

Morss, L. R.; Stanley, M.; Tatko, C.; Ebert, W. L.

1999-01-01

This paper reports the results of corrosion tests with monoliths of sodalite, binder glass, and glass-bonded sodalite, a ceramic waste form (CWF) that is being developed to immobilize radioactive electrorefiner salt used to condition spent sodium-bonded nuclear fuel. These tests were performed with dilute pH-buffered solutions in the pH range of 5-10 at temperatures of 70 and 90 C. The pH dependence of the forward dissolution rates of the CWF and its components have been determined. The pH dependence of the dissolution rates of sodalite, binder glass, and glass-bonded sodalite are similar to the pH dependence of dissolution rate of borosilicate nuclear waste glasses, with a negative pH dependence in the acidic region and a positive pH dependence in the basic region. Our results on the forward dissolution rates and their temperature and pH dependence will be used as components of a waste form degradation model to predict the long-term behavior of the CWF in a nuclear waste repository

Leaching of actinides from nuclear waste glass: French experience

International Nuclear Information System (INIS)

Vernaz, E.Y.; Godon, N.

1991-01-01

The activity concentration versus time of a typical LWR glass shows that after 300 years most of the activity is attributable to three actinides (Np, Pu and Am) and to 99 Tc. This activity decreases slowly, and some 50.000 years are necessary before the activity concentration drops to the level of the richest natural ores. This paper reviews the current state of knowledge concerning the kinetics of actinide release from glass subjected to aqueous leaching

Investigation of lead-iron-phosphate glass for SRP waste

International Nuclear Information System (INIS)

Jantzen, C.M.

1986-10-01

The search for a host solid for the immobilization of nuclear waste has focused on various vitreous waste forms. Recently, lead-iron-phosphate (LIP) glasses have been proposed for solidification of all types of HLLW. Investigation of this glass for vitrification of SRP waste demonstrated that the phosphate glass is incompatible with the current borosilicate glass technology. The durability of LIP glasses in deionized water was comparable to current borosilicate waste glass formulations, and the LIP glass has a low melt temperature. However, many of the defense waste constituents have low solubility in the phosphate melt, producing an inhomogeneous product. Also, the LIP melt is highly corrosive which prevents the use of current melter materials, in particular Inconel 690, and thus requires more exotic materials of construction such as platinum

Alteration of basaltic glass in Iceland as a natural analogue for nuclear waste glasses

International Nuclear Information System (INIS)

Crovisier, J.L.; Advocat, T.; Fritz, B.; Petit, J.C.

1989-01-01

The authors describe the longer term geochemical consequences of basaltic glass dissolution in fresh water at 0 0 C calculated with the computer code DISSOL. The clay minerals were represented by an ideal solid solution model (CISSFIT) able to describe variations in chemical composition of a clay phase in response to variations of the solution chemistry. The predicted mineral phases were iron hydroxides followed by kaolinite, TOT clays, chabazite and clinoptilolite. These results are in reasonably good agreement with experimental results and observations of altered subglacial hyaloclastites from Iceland. The formation of secondary products are mainly controlled by thermodynamic constraints. Kinetic effects, such as diffusion in the near glass surface are not important

57Fe Moessbauer effect in borosilicate glasses

International Nuclear Information System (INIS)

Music, S.

1989-01-01

The present study was carried out to elucidate the valence state of iron and its co-ordination in borosilicate glasses, which are being investigated as possible solidification matrices for the immobilization of a simulated nuclear waste. 57 Fe Mossbauer spectroscopy was used as the experimental technique. The chemical compositions of glass samples and the experimental conditions for the preparation of these samples are given. Iron in the form of haematite (α-Fe 2 O 3 ) was used as doping material. Details of the experimental procedure have previously been described. Isomer shifts are calculated relative to α-iron. The results indicate a strong dependence of the valency of the iron and its coordination on the chemical composition of the glass and the Fe 2 O 3 content. The method of preparing the glasses also influences the state of the iron in oxide glasses. (Author)

In situ study of glasses decomposition layer

International Nuclear Information System (INIS)

Zarembowitch-Deruelle, O.

1997-01-01

The aim of this work is to understand the involved mechanisms during the decomposition of glasses by water and the consequences on the morphology of the decomposition layer, in particular in the case of a nuclear glass: the R 7 T 7 . The chemical composition of this glass being very complicated, it is difficult to know the influence of the different elements on the decomposition kinetics and on the resulting morphology because several atoms have a same behaviour. Glasses with simplified composition (only 5 elements) have then been synthesized. The morphological and structural characteristics of these glasses have been given. They have then been decomposed by water. The leaching curves do not reflect the decomposition kinetics but the solubility of the different elements at every moment. The three steps of the leaching are: 1) de-alkalinization 2) lattice rearrangement 3) heavy elements solubilization. Two decomposition layer types have also been revealed according to the glass heavy elements rate. (O.M.)

Plutonium dioxide dissolution in glass

International Nuclear Information System (INIS)

Vienna, J.D.; Alexander, D.L.; Li, Hong

1996-09-01

In the aftermath of the Cold War, the U.S. Department of Energy's (DOE) Office of Fissile Materials Disposition (OFMD) is charged with providing technical support for evaluation of disposition options for excess fissile materials manufactured for the nation's defense. One option being considered for the disposition of excess plutonium (Pu) is immobilization by vitrification. The vitrification option entails immobilizing Pu in a host glass and waste package that are criticality-safe (immune to nuclear criticality), proliferation-resistant, and environmentally acceptable for long-term storage or disposal. To prove the technical and economic feasibility of candidate vitrification options it is necessary to demonstrate that PuO 2 feedstock can be dissolved in glass in sufficient quantity. The OFMD immobilization program has set a Pu solubility goal of 10 wt% in glass. The life cycle cost of the vitrification options are strongly influenced by the rate at which PUO 2 dissolves in glass. The total number of process lines needed for vitrification of 50 t of Pu in 10 years is directly dependent upon the time required for Pu dissolution in glass. The objective of this joint Pacific Northwest National Laboratory (PNNL) - Savannah River Technology Center (SRTC) study was to demonstrate a high Pu solubility in glass and to identify on a rough scale the time required for Pu dissolution in the glass. This study was conducted using a lanthanide borosilicate (LaBS) glass composition designed at the SRTC for the vitrification of actinides

Plutonium dioxide dissolution in glass

Energy Technology Data Exchange (ETDEWEB)

Vienna, J.D.; Alexander, D.L.; Li, Hong [and others

1996-09-01

In the aftermath of the Cold War, the U.S. Department of Energy`s (DOE) Office of Fissile Materials Disposition (OFMD) is charged with providing technical support for evaluation of disposition options for excess fissile materials manufactured for the nation`s defense. One option being considered for the disposition of excess plutonium (Pu) is immobilization by vitrification. The vitrification option entails immobilizing Pu in a host glass and waste package that are criticality-safe (immune to nuclear criticality), proliferation-resistant, and environmentally acceptable for long-term storage or disposal. To prove the technical and economic feasibility of candidate vitrification options it is necessary to demonstrate that PuO{sub 2} feedstock can be dissolved in glass in sufficient quantity. The OFMD immobilization program has set a Pu solubility goal of 10 wt% in glass. The life cycle cost of the vitrification options are strongly influenced by the rate at which PUO{sub 2} dissolves in glass. The total number of process lines needed for vitrification of 50 t of Pu in 10 years is directly dependent upon the time required for Pu dissolution in glass. The objective of this joint Pacific Northwest National Laboratory (PNNL) - Savannah River Technology Center (SRTC) study was to demonstrate a high Pu solubility in glass and to identify on a rough scale the time required for Pu dissolution in the glass. This study was conducted using a lanthanide borosilicate (LaBS) glass composition designed at the SRTC for the vitrification of actinides.

MoO3 incorporation in magnesium aluminosilicate glasses

International Nuclear Information System (INIS)

Tan, Shengheng; Ojovan, Michael I.; Hyatt, Neil C.; Hand, Russell J.

2015-01-01

Molybdate has a very low solubility in silicate and borosilicate glass systems and its excess presence in nuclear waste glass can cause the formation of a readily soluble “yellow phase”. In this study, the incorporation of molybdenum oxide (MoO 3 ) in a magnesium aluminosilicate glass system has been investigated. The prepared glasses show a higher than 90% molybdenum retention rate and up to 5.34 mol% (12.28 wt%) MoO 3 can be incorporated into these glasses without causing visible phase separation. The incorporation of MoO 3 increases glass density, decreases glass transition and crystallisation temperatures and intensifies Raman bands assigned to vibrations of MoO 4 2− units. When excess molybdate is added liquid–liquid phase separation and crystallisation occurs. The separated phase is spherical, 200–400 nm in diameter and randomly dispersed. Based on powder X-ray diffraction, Raman spectroscopy and transmission electron microscopy, the separated phase is identified as MgMoO 4

Remaining uncertainties in predicting long-term performance of nuclear waste glass from experiments

International Nuclear Information System (INIS)

Grambow, B.

1994-01-01

The current knowledge on the glass dissolution mechanism and the representation of glass dissolution concepts within overall repository performance assessment models are briefly summarized and uncertainties related to mechanism, radionuclide chemistry and parameters are discussed. Understanding of the major glass dissolution processes has been significantly increased in recent years. Long-term glass stability is related to the long-term maintenance of silica saturated conditions. The behavior of individual radionuclides in the presence of a dissolving glass has not been sufficiently and results do no yet allow meaningful predictions. Conserving long-term predictions of glass matrix dissolution as upper limit for radionuclide release can be made with sufficient confidence, however these estimations generally result in a situation where the barrier function of the glass is masked by the efficiency of the geologic barrier. Realistic long-term predictions may show that the borosilicate waste glass contributes to overall repository safety to a much larger extent than indicated by overconservatism. Today realistic predictions remain highly uncertain and much more research work is necessary. In particular, the long-term rate under silica saturated conditions needs to be understood and the behavior of individual radionuclides in the presence of a dissolving glass deserves more systematic investigations

Towards optimization of nuclear waste glass: Constraints, property models, and waste loading

International Nuclear Information System (INIS)

Hrma, P.

1994-04-01

Vitrification of both low- and high-level wastes from 177 tanks at Hanford poses a great challenge to glass makers, whose task is to formulate a system of glasses that are acceptable to the federal repository for disposal. The enormous quantity of the waste requires a glass product of the lowest possible volume. The incomplete knowledge of waste composition, its variability, and lack of an appropriate vitrification technology further complicates this difficult task. A simple relationship between the waste loading and the waste glass volume is presented and applied to the predominantly refractory (usually high-activity) and predominantly alkaline (usually low-activity) waste types. Three factors that limit waste loading are discussed, namely product acceptability, melter processing, and model validity. Glass formulation and optimization problems are identified and a broader approach to uncertainties is suggested

SETTLING OF SPINEL IN A HIGH-LEVEL WASTE GLASS MELTER

International Nuclear Information System (INIS)

Pavel Hrma; Pert Schill; Lubomir Nemec

2002-01-01

High-level nuclear waste is being vitrified, i.e., converted to a durable glass that can be stored in a safe repository for hundreds of thousands of years. Waste vitrification is accomplished in reactors called melters to which the waste is charged together with glass-forming additives. The mixture is electrically heated to a temperature as high as 1150 decrees C to create a melt that becomes glass on cooling

Surface layer effects on waste glass corrosion

International Nuclear Information System (INIS)

Feng, X.

1993-01-01

Water contact subjects waste glass to chemical attack that results in the formation of surface alteration layers. Two principal hypotheses have been advanced concerning the effect of surface alteration layers on continued glass corrosion: (1) they act as a mass transport barrier and (2) they influence the chemical affinity of the glass reaction. In general, transport barrier effects have been found to be less important than affinity effects in the corrosion of most high-level nuclear waste glasses. However, they can be important under some circumstances, for example, in a very alkaline solution, in leachants containing Mg ions, or under conditions where the matrix dissolution rate is very low. The latter suggests that physical barrier effect may affect the long-term glass dissolution rate. Surface layers influence glass reaction affinity through the effects of the altered glass and secondary phases on the solution chemistry. The reaction affinity may be controlled by various precipitates and crystalline phases, amorphous silica phases, gel layer, or all the components of the glass. The surface alteration layers influence radionuclide release mainly through colloid formation, crystalline phase incorporation, and gel layer retention. This paper reviews current understanding and uncertainties

Glass formulation development and offgas analysis of microwave melter powder samples

International Nuclear Information System (INIS)

Semones, G.B.; Hoffman, C.R.; Phillips, J.A.

1994-04-01

Production of nuclear materials for defense applications has resulted in the accumulation of vast amounts of nuclear waste. This contaminated waste is in a variety of forms that require subsequent reprocessing to isolate and encapsulate the nuclear (e.g., uranium, plutonium, strontium, cesium, and americium) and toxic (e.g., lead, chromium, and cadmium) constituents. The encapsulating material must possess good chemical and mechanical durability to resist leaching of the nuclear and toxic constituents into the environment during permanent storage at a waste repository. Glass is an ideal encapsulating material because its open structure allows the introduction of different waste forms and the final vitreous product possesses a high degree of chemical stability. Microwave heating and melting is a relatively new advancement in glass processing which uses microwave radiation to heat the glass formers to adequate temperatures for sintering or melting. An advantage to this technique is that it enables more rapid heating than traditional heating mechanisms. This decrease in cycle time may help to limit exposure to workers encapsulating radioactive and/or toxic waste

Structural analysis and thermal behavior of diopside-fluorapatite-wollastonite-based glasses and glass-ceramics.

Science.gov (United States)

Kansal, Ishu; Tulyaganov, Dilshat U; Goel, Ashutosh; Pascual, Maria J; Ferreira, José M F

2010-11-01

Glass-ceramics in the diopside (CaMgSi2O6)-fluorapatite (Ca5(PO4)3F)-wollastonite (CaSiO3) system are potential candidates for restorative dental and bone implant materials. The present study describes the influence of varying SiO2/CaO and CaF2/P2O5 molar ratio on the structure and thermal behavior of glass compositions in the CaO-MgO-SiO2-P2O5-Na2O-CaF2 system. The structural features and properties of the glasses were investigated by nuclear magnetic resonance (NMR), infrared spectroscopy, density measurements and dilatometry. Sintering and crystallization behavior of the glass powders were studied by hot-stage microscopy and differential thermal analysis, respectively. The microstructure and crystalline phase assemblage in the sintered glass powder compacts were studied under non-isothermal heating conditions at 825 °C. X-ray diffraction studies combined with the Rietveld-reference intensity ratio (R.I.R) method were employed to quantify the amount of amorphous and crystalline phases in the glass-ceramics, while scanning electron microscopy was used to shed some light on the microstructure of resultant glass-ceramics. An increase in CaO/SiO2 ratio degraded the sinterability of the glass powder compacts, resulting in the formation of akermanite as the major crystalline phase. On the other hand, an increase in P2O5/CaF2 ratio improved the sintering behavior of the glass-ceramics, while varying the amount of crystalline phases, i.e. diopside, fluorapatite and wollastonite. Copyright © 2010. Published by Elsevier Ltd.

Effect of composition and radiation on the Hertzian indentation behavior of nuclear waste glasses

International Nuclear Information System (INIS)

Matzke, H.; Kahl, L.; Routbort, J.L.; Saidl, J.

1983-01-01

The Hertzian indentation technique has been used to determine the fracture toughness, K/sub Ic/ of two borosilicate glasses developed to contain high-level nuclear waste. For the product VG 98/12, adding selected groups of fission products leaves K/sub Ic/ unchanged, but addition of Pb lowers K/sub Ic/ by approx. 20%. Radiation with 77 MeV α-particles to a dose of approx. 10 15 α/cm 2 increases K/sub Ic/ by approx. 75%. For the product SM 58 LW 11, the fracture toughness was measured on pieces taken from different parts of a large cylinder to investigate the effects of segregation phenomena and of partial crystallization and formation of small cristobalite inclusions which decrease K/sub Ic/ by approx. 25%

Characterization of high level nuclear waste glass samples following extended melter idling

Energy Technology Data Exchange (ETDEWEB)

Fox, Kevin M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Peeler, David K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kruger, Albert A. [USDOE Office of River Protection, Richland, WA (United States)

2015-06-16

The Savannah River Site Defense Waste Processing Facility (DWPF) melter was recently idled with glass remaining in the melt pool and riser for approximately three months. This situation presented a unique opportunity to collect and analyze glass samples since outages of this duration are uncommon. The objective of this study was to obtain insight into the potential for crystal formation in the glass resulting from an extended idling period. The results will be used to support development of a crystal-tolerant approach for operation of the high-level waste melter at the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Two glass pour stream samples were collected from DWPF when the melter was restarted after idling for three months. The samples did not contain crystallization that was detectible by X-ray diffraction. Electron microscopy identified occasional spinel and noble metal crystals of no practical significance. Occasional platinum particles were observed by microscopy as an artifact of the sample collection method. Reduction/oxidation measurements showed that the pour stream glasses were fully oxidized, which was expected after the extended idling period. Chemical analysis of the pour stream glasses revealed slight differences in the concentrations of some oxides relative to analyses of the melter feed composition prior to the idling period. While these differences may be within the analytical error of the laboratories, the trends indicate that there may have been some amount of volatility associated with some of the glass components, and that there may have been interaction of the glass with the refractory components of the melter. These changes in composition, although small, can be attributed to the idling of the melter for an extended period. The changes in glass composition resulted in a 70-100 °C increase in the predicted spinel liquidus temperature (TL) for the pour stream glass samples relative to the analysis of the melter feed prior to

Effects of alpha radiation on hardness and toughness of the borosilicate glass applied to radioactive wastes immobilization

International Nuclear Information System (INIS)

Prado, Miguel Oscar; Bernasconi, Norma B. Messi de; Bevilacqua, Arturo Miguel; Arribere, Maria Angelica; Heredia, Arturo D.; Sanfilippo, Miguel

1999-01-01

Borosilicate german glass SG7 samples, obtained by frit sintering, were irradiated with different fluences of thermal neutrons in the nucleus of a nuclear reactor. The nuclear reaction 10 B(n,α) 7 Li, where the 10 B isotope is one of the natural glass components, was used to generate alpha particles throughout the glass volume. The maximum alpha disintegration per unit volume achieved was equivalent to that accumulated in a borosilicate glass with nuclear wastes after 3.8 million years. Through Vickers indentations values for microhardness, stress for 50% fracture probability (Weibull statistics) and estimation of the toughness were obtained as a function of alpha radiation dose. Two counterbalanced effects were found: that due to the disorder created by the alpha particles in the glass and that due to the annealing during irradiation (temperature below 240 deg C). Considering the alpha radiation effect, glasses tend decrease Vickers hardness, and to increase thr 50% fracture probability stress with the dose increase. (author)

Use of molecular dynamics to model the structure of nuclear glasses

International Nuclear Information System (INIS)

Delaye, J.M.; Ghaleb, D.

1997-01-01

Born-Mayer-Huggins potentials (BMH) are generally used to model oxides glasses. In this article we show how they can be used, if they can be completed by three body terms, to reproduce correctly structures of alumino-borosilicate glasses. Taking into account the parameters adjusted for the basic matrix (SiO 2 , B 2 O 3 , Na 2 O, ZrO 2 , Al 2 O 3 ) of French light water reactor waste containment glass, we have analyzed equally the structure evolutions reproduced by BMH potentials according to the modification of the composition of the glass. The principal objective of this study being the appreciation of the capability of this potentials to simulate structure evolutions in the glass; this potentials representing neither the nature of chemical bonds nor the electronic state modifications that accompany changes of coordination numbers. Globally BHM potentials, despite their limited representativeness, reproduce some experimental observations. For example: - the gradual incorporation of the boron in the silicate network; - the shorter distances between network formers and non-bridging oxygen atoms; - the visualization of boron-enriched segregation zones in a high content boron composition without aluminum. (authors)

Millimeter-Wave Measurements of High Level and Low Level Activity Glass Melts

International Nuclear Information System (INIS)

Woskov, Paul

2005-01-01

EMSP supported research of millimeter-wave technology for nuclear waste glass melter monitoring has been very productive in establishing this field and showing great progress. This work has garnered significant recognition, winning an R and D 100 Award for viscosity monitoring, a Best Paper Award by the American Ceramic Society for nuclear waste glass monitoring, investment by the Glass Plus industry consortium to test this technology for glass fiber manufacture, investment by Savannah River Technology Center in purchasing key hardware components for additional tests, and Japanese initiated exchange visits between MIT and the vitrification facilities at Japanese Atomic Energy Research Institute (JAERI) in Tokai to review this technology. There are also potentially important spin offs to other areas including nuclear and fossil fuel power production, and National Institute of Health sponsored research as indicated below. Consequently, this work has the potential of becoming a major inter nationally recognized EMSP success story. A summary of the main accomplishments follows. The readers are referred to the cited reference publications for more details, many of which were EMSP supported by this work

The chemistry of copper chalcogenides in waste glasses

International Nuclear Information System (INIS)

Schreiber, H.D.; Lambert, H.W.

1994-01-01

The solubilities of copper chalcogenides (CuS, CuSe, CuTe) were measured in a glass melt which is representative of those proposed for nuclear waste immobilization and circuit board vitrification. CuTe is more soluble than CuS and CuSe in the glass melt under relatively oxidizing conditions. However, the solubilities of all the copper chalcogenides in the glass melt are virtually identical at reducing conditions, probably a result of the redox-controlled solubility of copper metal in all cases. The redox chemistry of a glass melt coexisting with an immiscible copper chalcogenide depends primarily on the prevailing oxygen fugacity, not on the identity of the chalcogenide. The target concentration of less than 0.3 to 0.5 wt% copper in the waste glass should eliminate the precipitation of copper chalcogenides during processing

RHENIUM SOLUBILITY IN BOROSILICATE NUCLEAR WASTE GLASS IMPLICATIONS FOR THE PROCESSING AND IMMOBILIZATION OF TECHNETIUM-99 (AND SUPPORTING INFORMATION WITH GRAPHICAL ABSTRACT)

Energy Technology Data Exchange (ETDEWEB)

AA KRUGER; A GOEL; CP RODRIGUEZ; JS MCCLOY; MJ SCHWEIGER; WW LUKENS; JR, BJ RILEY; D KIM; M LIEZERS; P HRMA

2012-08-13

The immobilization of 99Tc in a suitable host matrix has proved a challenging task for researchers in the nuclear waste community around the world. At the Hanford site in Washington State in the U.S., the total amount of 99Tc in low-activity waste (LAW) is {approx} 1,300 kg and the current strategy is to immobilize the 99Tc in borosilicate glass with vitrification. In this context, the present article reports on the solubility and retention of rhenium, a nonradioactive surrogate for 99Tc, in a LAW sodium borosilicate glass. Due to the radioactive nature of technetium, rhenium was chosen as a simulant because of previously established similarities in ionic radii and other chemical aspects. The glasses containing target Re concentrations varying from 0 to10,000 ppm by mass were synthesized in vacuum-sealed quartz ampoules to minimize the loss of Re by volatilization during melting at 1000 DC. The rhenium was found to be present predominantly as Re7 + in all the glasses as observed by X-ray absorption near-edge structure (XANES). The solubility of Re in borosilicate glasses was determined to be {approx}3,000 ppm (by mass) using inductively coupled plasma-optical emission spectroscopy (ICP-OES). At higher rhenium concentrations, some additional material was retained in the glasses in the form of alkali perrhenate crystalline inclusions detected by X-ray diffraction (XRD) and laser ablation-ICP mass spectrometry (LA-ICP-MS). Assuming justifiably substantial similarities between Re7 + and Tc 7+ behavior in this glass system, these results implied that the processing and immobilization of 99Tc from radioactive wastes should not be limited by the solubility of 99Tc in borosilicate LAW glasses.

Fabrication and characterization of MCC approved testing material - ATM-8 glass

International Nuclear Information System (INIS)

Wald, J.W.

1985-10-01

The Materials Characterization Center (MCC) Approved Testing Material ATM-8 is a borosilicate glass that incorporates elements typical of high-level waste (HLW) resulting from the reprocessing of commercial nuclear reactor fuel. Its composition is based upon the simulated HLW glass type 76-68 (Mendel, J.E. et al., 1977, Annual Report of the Characteristics of High-Level Waste Glasses, BNWL-2252, Pacific Northwest Laboratory, Richland, Washington), to which depleted uranium, technetium-99, neptunium-237 and plutonium-239 have been added at moderate to low levels. The glass was requested by the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. It was produced by the MCC at the Pacific Northwest Laboratory (PNL) operated for the Department of Energy (DOE) by Battelle Memorial Institute. ATM-8 glass was produced in April of 1984, and is the second in a series of testing materials for NNWSI. This report discusses its fabrication (starting materials, batch and glass preparation, measurement and testing equipment, other equipment, procedures, identification system and materials availability and storage, and characterization (bulk density) measurements, chemical analysis, microscopic examination, and x-ray diffraction analysis. 4 refs., 2 figs., 10 tabs

Thermal property of holmium doped lithium lead borate glasses

Science.gov (United States)

Usharani, V. L.; Eraiah, B.

2018-04-01

The new glass system of holmium doped lithium lead borate glasses were prepared by conventional melt quenching technique. The thermal stability of the different compositions of Ho3+ ions doped lithium lead borate glasses were studied by using TG-DTA. The Tg values are ranging from 439 to 444 °C with respect to the holmium concentration. Physical parameters like polaron radius(rp), inter-nuclear distance (ri), field strength (F) and polarizability (αm) of oxide ions were calculated using appropriate formulae.

Localized chemistry of 99Tc in simulated low activity waste glass

Science.gov (United States)

Weaver, Jamie L.

A priority of the United States Department of Energy (DOE) is to dispose of the nuclear waste accumulated in the underground tanks at the Hanford Nuclear Reservation in Richland, WA. Incorporation and stabilization of technetium (99Tc) from these tanks into vitrified waste forms is a concern to the waste glass community and DOE due to 99Tc's long half-life ( 2.13˙105 y), and its high mobility in the subsurface environment under oxidizing conditions. Working in collaboration with researchers at Pacific Northwest National Laboratory (PNNL) and other national laboratories, plans were formulated to obtain first-of-a-kind chemical structure determination of poorly understood and environmentally relevant technetium compounds that relate to the chemistry of the Tc in nuclear waste glasses. Knowledge of the structure and spectral signature of these compounds aid in refining the understanding of 99Tc incorporation into and release from oxide based waste glass. In this research a first-of-its kind mechanism for the behavior of 99Tc during vitrification is presented, and the structural role of Tc(VII) and (IV) in borosilicate waste glasses is readdressed.

Laser glass: a key material in the search for fusion energy

International Nuclear Information System (INIS)

Campbell, J H

1999-01-01

Nuclear fusion is the energy source that powers the sun. For more than four decades man has sought to develop this essentially inexhaustible, clean power source for use on earth. Unfortunately the conditions needed to initiate fusion are daunting; the nuclear fuel, consisting of isotopes of hydrogen, must be heated to temperatures in excess of 100,000,000 C and maintained at that temperature long enough for the nuclear fuel to ignite and burn. Lasers are being used as one of the tools to achieve these conditions. The best lasers for this work are those that derive their energy from a unique set of optical glasses called laser glasses. The work to develop, manufacture and test these glasses has involved a partnership between university and industry that has spanned more than 25 years. During this time lasers used in fusion development have grown from small systems that could fit on the top of a table to systems currently under construction that are approximately the size of a municipal sports stadium. A brief historical and anecdotal account of the development of laser glasses for fusion energy research applications is the subject of the presentation

Immobilization of high activity nuclear wastes in sintered glass. Fabrication of blocks at semi-industrial scale by hot pressing technique

International Nuclear Information System (INIS)

Russo, D.O.; Messi, N.B.; Riquelme, R.; Sterba, M.E.; Audero, M.A.

1990-01-01

The sintering process under glass pressure has been studied as an alternative of melting with the aim of obtaining a monolytic material apt to preserve the high activity nuclear wastes. Different properties of the products obtained have been evaluated where the material is selected on the basis of the results attained. The purpose of this work is the equipment development and the process adjusting for the blocks obtainment. (Author) [es

Critical review of glass performance modeling

International Nuclear Information System (INIS)

Bourcier, W.L.

1994-07-01

Borosilicate glass is to be used for permanent disposal of high-level nuclear waste in a geologic repository. Mechanistic chemical models are used to predict the rate at which radionuclides will be released from the glass under repository conditions. The most successful and useful of these models link reaction path geochemical modeling programs with a glass dissolution rate law that is consistent with transition state theory. These models have been used to simulate several types of short-term laboratory tests of glass dissolution and to predict the long-term performance of the glass in a repository. Although mechanistically based, the current models are limited by a lack of unambiguous experimental support for some of their assumptions. The most severe problem of this type is the lack of an existing validated mechanism that controls long-term glass dissolution rates. Current models can be improved by performing carefully designed experiments and using the experimental results to validate the rate-controlling mechanisms implicit in the models. These models should be supported with long-term experiments to be used for model validation. The mechanistic basis of the models should be explored by using modern molecular simulations such as molecular orbital and molecular dynamics to investigate both the glass structure and its dissolution process

Critical review of glass performance modeling

Energy Technology Data Exchange (ETDEWEB)

Bourcier, W.L. [Lawrence Livermore National Lab., CA (United States)

1994-07-01

Borosilicate glass is to be used for permanent disposal of high-level nuclear waste in a geologic repository. Mechanistic chemical models are used to predict the rate at which radionuclides will be released from the glass under repository conditions. The most successful and useful of these models link reaction path geochemical modeling programs with a glass dissolution rate law that is consistent with transition state theory. These models have been used to simulate several types of short-term laboratory tests of glass dissolution and to predict the long-term performance of the glass in a repository. Although mechanistically based, the current models are limited by a lack of unambiguous experimental support for some of their assumptions. The most severe problem of this type is the lack of an existing validated mechanism that controls long-term glass dissolution rates. Current models can be improved by performing carefully designed experiments and using the experimental results to validate the rate-controlling mechanisms implicit in the models. These models should be supported with long-term experiments to be used for model validation. The mechanistic basis of the models should be explored by using modern molecular simulations such as molecular orbital and molecular dynamics to investigate both the glass structure and its dissolution process.

Growth of hydrated gel layers in nuclear waste glasses

International Nuclear Information System (INIS)

Sullivan, T.M.; Machiels, A.J.

1984-01-01

The hydration kinetics of waste glasses in contact with an aqueous solution has been studied by using three different approaches. Emphasis has been placed on modeling processes in the transition zone defined as the region in which the nature of the glass changes from the original dry glass to an open hydrated structure. The first model relies on concentration-dependent diffusion coefficients to obtain a transition zone in which the ions mobility is extremely low compared to that in the gel layer. In the second model, the transition zone and hydrated layer are treated as distinct phases and it is assumed that ion exchange at their common boundary is the rate-controlling process. The third model treats the transition zone as a thin film of constant thickness and low diffusivity. In the absence of appreciable network dissolution, all three models indicate that growth of the gel layer becomes eventually proportional to the square root of time; however, as long as processes in the transition zone are rate controlling, growth is linearly proportional to time

An introduction to nuclear waste immobilisation

International Nuclear Information System (INIS)

Ojovan, M.I.; Lee, W.E.

2005-08-01

Safety and environmental impact is of uppermost concern when dealing with the movement and storage of nuclear waste. The 20 chapters in this book cover all important aspects of immobilisation, from nuclear decay, to regulations, to new technologies and methods. Significant focus is given to the analysis of the various matrices used in transport: cement, bitumen and glass, with the greatest attention being given to glass. The last chapter concentrates on the performance assessment of each matrix, and on new developments of ceramics and glass composite materials, thermochemical methods and in-situ metal matrix immobilisation. The book thoroughly covers all issues surrounding nuclear waste: from where to locate nuclear waste in the environment, through nuclear waste generation and sources, treatment schemes and technologies, immobilisation technologies and waste forms, disposal and long term behaviour. Particular attention is paid to internationally approved and worldwide-applied approaches and technologies

An alternative host matrix based on iron phosphate glasses for the vitrification of specialized nuclear waste forms. Annual progress report, September 15, 1996 - September 14, 1997

International Nuclear Information System (INIS)

Day, D.E.; Marasinghe, K.; Ray, C.S.

1997-01-01

'Objectives of this project are to: (1) investigate the glass composition and processing conditions that yield optimum properties for iron phosphate glasses for vitrifying radioactive waste, (2) determine the atomic structure of iron phosphate glasses and the structure-property relationships, (3) determine how the physical and structural properties of iron phosphate glasses are affected by the addition of simulated high level nuclear waste components, and (4) investigate the process and products of devitrification of iron phosphate waste forms. The glass forming ability of about 125 iron phosphate melts has been investigated in different oxidizing to reducing atmospheres using various iron oxide raw materials such as Fe 2 O 3 , FeO, Fe 3 O 4 , and FeC 2 O 4 2H 2 O. The chemical durability, redox equilibria between Fe(II) and Fe(III), crystallization behavior and structural features for these glasses and their crystalline forms have been investigated using a variety of techniques including Mossbauer spectroscopy, X-ray absorption spectroscopy (XAS), X-ray photoelectron spectroscopy (XPS), Extended x-ray absorption fine structure (EXAFS) and X-ray absorption near edge structure (XANES) analysis, differential thermal and thermogravimetric analysis (DTA/TGA), and X-ray and neutron diffraction.'

Platinoids and molybdenum in nuclear waste containment glasses: a structural study; Les platinoides et le molybdene dans des verres d'interet nucleaires: etude structurale

Energy Technology Data Exchange (ETDEWEB)

Le Grand, M [CEA/VALRHO - site de Marcoule, Dept. de Recherche en Retraitement et en Vitrification (DRRV), 30 - Marcoule (France); [Paris-7 Univ., 75 (France)

2000-07-01

This work deals with the structure of borosilicate nuclear glasses and with some relationships between structure and macroscopic properties. Two types of elements which may disturb the industrial process - platinoids (Ru and Pd) and molybdenum - are central to this work. Platinoids induce weak modifications on the structure of the glass, causing a depolymerization of the glassy network, an increase of the {sup [3]}B/{sup [4]}B ratio and a modification of the medium range order around Si between 3.3 and 4.5 angstrom. The modifications of viscosity and density induced by platinoids in the glass are not due to the structural effect of the platinoids. The increase of viscosity is attributed to needle shaped RuO{sub 2}. It can be moderated by imposing reducing conditions during the elaboration of the glass. The slight difference between experimental and calculated densities is due to the increase of the volume percentage of bubbles in the glass with increasing platinoid content. Mo is either present in the glass as molybdic groupings, or mobilized in chemically complex molybdic crystalline phases. The chemical composition and mineralogy of these phases has been obtained using electronic microprobe data and XRD with Rietveld analysis. The distribution of the different elements between the crystalline phases and the glass is strongly influenced by the structural role of the various cations in the glass. The Mo present in the glass appears as MoO{sub 4} tetrahedra, independent of the borosilicate network. The formation of the crystalline phases can be explained by the existence of a precursor in which the MoO{sub 4} tetrahedra are concentrated in rich alkali and earth-alkali bearing areas of the glass. (author)

New glass material oxidation and dissolution system facility: Direct conversion of surplus fissile materials, spent nuclear fuel, and other material to high-level-waste glass. Storage and disposition of weapons-usable fissile materials programmatic environmental impact statement data report: Predecisional draft

International Nuclear Information System (INIS)

Forsberg, C.W.; Elam, K.R.; Reich, W.J.

1995-01-01

With the end of the Cold War, countries have excess plutonium and other materials from the reductions in inventories of nuclear weapons. It has been recommended that these surplus fissile materials (SFMs) be processed so that they are no more accessible than plutonium in spent nuclear fuel (SNF). This SNF standard, if adopted worldwide, would prevent rapid recovery of SFMs for the manufacture of nuclear weapons. This report provides for the PEIS the necessary input data on a new method for the disposition of SFMs: the simultaneous conversion of SFMs, SNF, and other highly radioactive materials into high-level-waste (HLW) glass. The SFMs include plutonium, neptunium, americium, and 233 U. The primary SFM is plutonium. The preferred SNF is degraded SNF, which may require processing before it can be accepted by a geological repository for disposal. The primary form of this SNF is Hanford-N SNF with preirradiation uranium enrichments between 0.95 and 1.08%. The final product is a plutonium, low-enriched-uranium, HLW, borosilicate glass for disposition in a geological repository. The proposed conversion process is the Glass Material Oxidation and Dissolution System (GMODS), which is a new process. The initial analysis of the GMODS process indicates that a MODS facility for this application would be similar in size and environmental impact to the Defense Waste Processing Facility (DWPF) at the Savannah River Site. Because of this, the detailed information available on DWPF was used as the basis for much of the GMODS input into the SFMs PEIS

Spent fuel from nuclear research reactors immobilized in sintered glass

International Nuclear Information System (INIS)

Mateos, P.; Russo, D.O.; Rodriguez, D.; Heredia, A.; Sanfilippo, M.; Sterba, M.

2002-01-01

Different kinds of glasses, borosilicates, Iron borosilicates and Iron phosphates, were tested in order to determine its capability to immobilize calcined uranium silicide in a sintering process. Iron phosphate glass developed in our laboratory showed the best results in SEM analysis. Also its gravimetric leaching rate is less than 0.45 g.m -2 .day -1 for 7 and 10% loading which is lower than any previously studied for us. (author)

Minor component study for simulated high-level nuclear waste glasses (Draft)

International Nuclear Information System (INIS)

Li, H.; Langowskim, M.H.; Hrma, P.R.; Schweiger, M.J.; Vienna, J.D.; Smith, D.E.

1996-02-01

Hanford Site single-shell tank (SSI) and double-shell tank (DSI) wastes are planned to be separated into low activity (or low-level waste, LLW) and high activity (or high-level waste, HLW) fractions, and to be vitrified for disposal. Formulation of HLW glass must comply with glass processibility and durability requirements, including constraints on melt viscosity, electrical conductivity, liquidus temperature, tendency for phase segregation on the molten glass surface, and chemical durability of the final waste form. A wide variety of HLW compositions are expected to be vitrified. In addition these wastes will likely vary in composition from current estimates. High concentrations of certain troublesome components, such as sulfate, phosphate, and chrome, raise concerns about their potential hinderance to the waste vitrification process. For example, phosphate segregation in the cold cap (the layer of feed on top of the glass melt) in a Joule-heated melter may inhibit the melting process (Bunnell, 1988). This has been reported during a pilot-scale ceramic melter run, PSCM-19, (Perez, 1985). Molten salt segregation of either sulfate or chromate is also hazardous to the waste vitrification process. Excessive (Cr, Fe, Mn, Ni) spinel crystal formation in molten glass can also be detrimental to melter operation

Glass packages in interim storage

International Nuclear Information System (INIS)

Jacquet-Francillon, N.

1994-10-01

This report summarize the current state of knowledge concerning the behavior of type C waste packages consisting of vitrified high-level solutions produced by reprocessing spent fuel. The composition and the physical and chemical properties of the feed solutions are reviewed, and the vitrification process is described. Sodium alumino-borosilicate glass compositions are generally employed - the glass used at la Hague for LWR fuel solutions, for example, contains 45 % SiO 2 . The major physical, chemical, mechanical and thermal properties of the glass are reviewed. In order to allow their thermal power to diminish, the 3630 glass packages produced (as of January 1993) in the vitrification facilities at Marcoule and La Hague are placed in interim storage for several decades. The actual interim storage period has not been defined, as it is closely related to the concept and organization selected for the final destination of the packages: a geological repository. The glass behavior under irradiation is described. Considerable basic and applied research has been conducted to assess the aqueous leaching behavior of nuclear containment glass. The effects of various repository parameters (temperature, flow rate, nature of the environmental materials) have been investigated. The experimental findings have been used to specify a model describing the kinetics of aqueous corrosion of the glass. More generally all the ''source term'' models developed in France by the CEA or by ANDRA are summarized. (author). 152 refs., 33 figs

Preliminary results of durability testing with borosilicate glass compositions

International Nuclear Information System (INIS)

Adel-Hadadi, M.; Adiga, R.; Barkatt, Aa.

1987-01-01

This is a report on the first year of research conducted at the Vitreous State Laboratory of the Catholic University of America in support of the West Valley Demonstration Project. One objective is the vitrification of liquid waste generated by previous nuclear fuel reprocessing. This work has been directed principally at the problem of glass composition optimization. This has necessitated the development of a coordinated program of glass production, durability measurements, and processability assessment. A small-scale continuous melter has been constructed for melting uranium and thorium containing glasses and for studying glass processing characteristics. Glass viscosities have been measured over a range of temperatures. A large number of glasses have also been produced in small crucible melts. Glass durability has been assessed using four types of leach tests: MCC-3, MCC-1, IAEA/ISO, and pulsed-flow tests. Extensive data from these tests are reported. The data have led to the design of very durable glasses (comparable to the Savannah River Laboratory Defense Waste Reference Glass) which have the requisite waste loading and processing characteristics. 14 refs., 4 figs., 77 tabs

Accelerated Leach Testing of GLASS: ALTGLASS Version 3.0

Energy Technology Data Exchange (ETDEWEB)

Trivelpiece, Cory L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Jantzen, Carol M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Crawford, Charles L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-12-31

The Accelerated Leach Testing of GLASS (ALTGLASS) database is a collection of data from short- and long-term product consistency tests (PCT, ASTM C1285 A and B) on high level waste (HLW) as well as low activity waste (LAW) glasses. The database provides both U.S. and international researchers with an archive of experimental data for the purpose of studying, modeling, or validating existing models of nuclear waste glass corrosion. The ALTGLASS database is maintained and updated by researchers at the Savannah River National Laboratory (SRNL). This newest version, ALTGLASS Version 3.0, has been updated with an additional 503 rows of data representing PCT results from corrosion experiments conducted in the United States by the Savannah River National Laboratory, Pacific Northwest National Laboratory, Argonne National Laboratory, and the Vitreous State Laboratory (SRNL, PNNL, ANL, VSL, respectively) as well as the National Nuclear Laboratory (NNL) in the United Kingdom.

Accelerated Leach Testing of GLASS: ALTGLASS Version 3.0

International Nuclear Information System (INIS)

Trivelpiece, Cory L.; Jantzen, Carol M.; Crawford, Charles L.

2016-01-01

The Accelerated Leach Testing of GLASS (ALTGLASS) database is a collection of data from short- and long-term product consistency tests (PCT, ASTM C1285 A and B) on high level waste (HLW) as well as low activity waste (LAW) glasses. The database provides both U.S. and international researchers with an archive of experimental data for the purpose of studying, modeling, or validating existing models of nuclear waste glass corrosion. The ALTGLASS database is maintained and updated by researchers at the Savannah River National Laboratory (SRNL). This newest version, ALTGLASS Version 3.0, has been updated with an additional 503 rows of data representing PCT results from corrosion experiments conducted in the United States by the Savannah River National Laboratory, Pacific Northwest National Laboratory, Argonne National Laboratory, and the Vitreous State Laboratory (SRNL, PNNL, ANL, VSL, respectively) as well as the National Nuclear Laboratory (NNL) in the United Kingdom.

Corrosion mechanisms of containment glasses for fission products

International Nuclear Information System (INIS)

Nogues, J.L.

1984-01-01

After a review of nuclear energy production and waste vitrification principles, the aqueous corrosion mechanisms of the containment glasses and the various parameters affecting the corrosion are studied: effects of glass composition, temperature, lixiviation agent pH, lixiviation duration and mode. Conventional mass loss measurement and solution analyses are coupled to sophisticated surface analysis techniques. The hydrolyzed layer formation and the solubility limits are discussed. 87 figs., 30 tabs., 144 refs

Laboratory work in support of West Valley glass development

International Nuclear Information System (INIS)

Bunnell, L.R.

1988-05-01

Over the past six years, Pacific Northwest Laboratory (PNL) has conducted several studies in support of waste glass composition development and testing of glass compositions suitable for immobilizing the nuclear wastes stored at West Valley, New York. As a result of pilot-scale testing conducted by PNL, the glass composition was changed from that originally recommended in response to changes in the waste stream, and several processing-related problems were discovered. These problems were solved, or sufficiently addressed to determine their likely effect on the glass melting operations to be conducted at West Valley. This report describes the development of the waste glass composition, WV-205, and discusses solutions to processing problems such as foaming and insoluble sludges, as well as other issues such as effects of feed variations on processing of the resulting glass. An evaluation of the WV-205 glass from a repository perspective is included in the appendix to this report

Liquidus temperature and chemical durability of selected glasses to immobilize rare earth oxides waste

Energy Technology Data Exchange (ETDEWEB)

Mohd Fadzil, Syazwani, E-mail: mfsyazwani86@postech.ac.kr [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology, 790784 Pohang, Gyeongbuk (Korea, Republic of); School of Applied Physics, Faculty of Science and Technology, The National University of Malaysia, 43650 Bandar Baru Bangi, Selangor (Malaysia); Hrma, Pavel [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology, 790784 Pohang, Gyeongbuk (Korea, Republic of); Pacific Northwest National Laboratory, P.O. Box 999, Richland, WA (United States); Schweiger, Michael J.; Riley, Brian J. [Pacific Northwest National Laboratory, P.O. Box 999, Richland, WA (United States)

2015-10-15

Pyroprocessing is are processing method for managing and reusing used nuclear fuel (UNF) by dissolving it in an electrorefiner with a molten alkali or alkaline earth chloride salt mixture while avoiding wet reprocessing. Pyroprocessing UNF with a LiCl–KCl eutectic salt releases the fission products from the fuel and generates a variety of metallic and salt-based species, including rare earth (RE) chlorides. If the RE-chlorides are converted to oxides, borosilicate glass is a prime candidate for their immobilization because of its durability and ability to dissolve almost any RE waste component into the glass matrix at high loadings. Crystallization that occurs in waste glasses as the waste loading increases may complicate glass processing and affect the product quality. This work compares three types of borosilicate glasses in terms of liquidus temperature (T{sub L}): the International Simple Glass designed by the International Working Group, sodium borosilicate glass developed by Korea Hydro and Nuclear Power, and the lanthanide aluminoborosilicate (LABS) glass established in the United States. The LABS glass allows the highest waste loadings (over 50 mass% RE{sub 2}O{sub 3}) while possessing an acceptable chemical durability. - Highlights: • We investigated crystallization in borosilicate glasses containing rare earth oxides. • New crystallinity and durability data are shown for glasses proposed in the literature. • Both liquidus temperature and chemical durability increased as the waste loading increased.

Solid state nuclear magnetic resonance: investigating the spins of nuclear related materials

International Nuclear Information System (INIS)

Charpentier, Th.

2007-10-01

The author reviews his successive research works: his research thesis work on the Multiple Quantum Magic Angle Spinning (MQMAS) which is a quadric-polar nucleus multi-quanta correlation spectroscopy method, the modelling of NMR spectra of disordered materials, the application to materials of interest for the nuclear industry (notably the glasses used for nuclear waste containment). He presents the various research projects in which he is involved: storing glasses, nuclear magnetic resonance in paramagnetism, solid hydrogen storing matrices, methodological and instrument developments in high magnetic field and high resolution solid NMR, long range distance measurement by solid state Tritium NMR (observing the structure and dynamics of biological complex systems at work)

Investigation of migration an actinide imitator (Ce) in synthesis glass ceramics with a nuclear microscope

International Nuclear Information System (INIS)

Bondarenko, V.N.; Goncharov, A.V.; Odejchuk, M.A.; Pistryak, V.M.; Pilipenko, A.V.; Sank, S.Yu.; Sukhostaeva, V.I.; Shevstiev, A.P.

2005-01-01

Full text: The generally accepted conception of radioactive waste (RAW) safe disposal, so-called 'multibarrier strategy', assumes construction of several protective barriers in a geologic burial. One of these barriers is is a durable corrosion-resistant container of RAW. The glass ceramics (syntetic alumina silicate) is considered as promising material for such containers. An important parameter for long term reduction of behavior for such materials is rate of radionuclide emanation from the container. As an element whose migration in solids is analogous as that of actinide, one generally uses Cerium. For investigation of transuranium element migration Ce is used as an analogue. Accordingly, CeO 2 may serve as an imitator of transuranium oxides. In the present work Ce migration in glass ceramics has been investigated with a nuclear microprobe which is more sensitive instrument than an electron microscope. The glass ceramic material was prepared from mixture of materials powders (70 % mas. of red granite and 30% mas. of kaolin) by sintering under pressure. The hot pressing was conducted at 1050 deg. temperature and 40 MPa pressure. The endurance time continued 15 min. Manufactured alumina silicate samples were in the form of a pellet. The material density was 2.5-2.7 g/cm 3 . Pellets of imitator substance were made of CeO 2 powder by the same method. After polishing, aluminosilicate and imitator pellets were pressed together and annealed in the temperature range from 600 to 750 deg. C in vacuum during 10 hours. As a result of annealing, Ce penetration occurred thought the interface into the depth of the alumina silicate pellet. After that the pellets were separated, and the alumina silicate pellet was cleaved in the direction transversal to the former interface. The surface of cleavage of alumina silicate pellet was studied with a nuclear microprobe. For measurement of Ce diffusion distribution in alumina silicate samples we applied a nuclear microprobe on the basis

Fabrication and characterization of MCC approved testing material - ATM-12 glass

International Nuclear Information System (INIS)

Wald, J.W.

1985-10-01

The Materials Characterization Center (MCC) Approved Testing Material ATM-12 is a borosilicate glass that incorporates elements typical of high-level waste (HLW) resulting from the reprocessing of commercial nuclear reactor fuels. The composition has been adjusted to match that predicted for HLW type 76-68 glass at an age of 300 y. Radioactive constituents contained in this glass include depleted uranium, 99 Tc, 237 Np, 239 Pu, and 241 Am. The glass was produced by the MCC at the Pacific Northwest Laboratory (PNL). ATM-12 glass ws produced from July to November of 1984 at the request of the Nevada Nuclear Waste Site Investigations (NNWSI) Program and is the third in a series of glasses produced for NNWSI. Most of the glass produced was in the form of cast bars; special castings and crushed material were also produced. Three kilograms of ATM-12 glass were produced from a feedstock melted in a nitrogen-atmosphere glove box at 1150 0 C in a platinum crucible, and formed into stress-annealed rectangular bars and the special casting shapes requested by NNWSI. Bars of ATM-12 were nominally 1.9 x 1.9 x 10 cm, with an average mass of 111 g each. Nineteen bars and 37 special castings were made. ATM-12 glass has been provided to the NNWSI Program, in the form of bars, crushed powder and special castings. As of August 1985 approximately 590 g of ATM-12 is available for distribution. Requests for materials or services related to this glass should be directed to the Materials Characterization Center Program Office, PNL

Encapsulation of TRISO particle fuel in durable soda-lime-silicate glasses

International Nuclear Information System (INIS)

Heath, Paul G.; Corkhill, Claire L.; Stennett, Martin C.; Hand, Russell J.; Meyer, Willem C.H.M.; Hyatt, Neil C.

2013-01-01

Tri-Structural Isotropic (TRISO) coated particle-fuel is a key component in designs for future high temperature nuclear reactors. This study investigated the suitability of three soda lime silicate glass compositions, for the encapsulation of simulant TRISO particle fuel. A cold press and sinter (CPS) methodology was employed to produce TRISO particle–glass composites. Composites produced were determined to have an aqueous durability, fracture toughness and Vickers’ hardness comparable to glasses currently employed for the disposal of high level nuclear wastes. Sintering at 700 °C for 30 min was found to remove all interconnected porosity from the composite bodies and oxidation of the outer pyrolytic carbon layer during sintering was prevented by processing under a 5% H 2 /N 2 atmosphere. However, the outer pyrolytic carbon layer was not effectively wetted by the encapsulating glass matrix. The aqueous durability of the TRISO particle–glass composites was investigated using PCT and MCC-1 tests combined with geochemical modelling. It was found that durability was dependent on silicate and calcium solution saturation. This study provides significant advancements in the preparation of TRISO particle encapsulant waste forms. The potential for the use of non-borosilicate sintered glass composites for TRISO particle encapsulation has been confirmed, although further refinements are required

Encapsulation of TRISO particle fuel in durable soda-lime-silicate glasses

Energy Technology Data Exchange (ETDEWEB)

Heath, Paul G.; Corkhill, Claire L.; Stennett, Martin C.; Hand, Russell J. [Immobilisation Science Laboratory, Department of Materials Science and Engineering, Robert Hadfield Building, University of Sheffield, Sheffield S1 3JD (United Kingdom); Meyer, Willem C.H.M. [Necsa, South African Nuclear Energy Corporation, PO Box 582, Pretoria, Gauteng (South Africa); Hyatt, Neil C., E-mail: n.c.hyatt@sheffield.ac.uk [Immobilisation Science Laboratory, Department of Materials Science and Engineering, Robert Hadfield Building, University of Sheffield, Sheffield S1 3JD (United Kingdom)

2013-05-15

Tri-Structural Isotropic (TRISO) coated particle-fuel is a key component in designs for future high temperature nuclear reactors. This study investigated the suitability of three soda lime silicate glass compositions, for the encapsulation of simulant TRISO particle fuel. A cold press and sinter (CPS) methodology was employed to produce TRISO particle–glass composites. Composites produced were determined to have an aqueous durability, fracture toughness and Vickers’ hardness comparable to glasses currently employed for the disposal of high level nuclear wastes. Sintering at 700 °C for 30 min was found to remove all interconnected porosity from the composite bodies and oxidation of the outer pyrolytic carbon layer during sintering was prevented by processing under a 5% H{sub 2}/N{sub 2} atmosphere. However, the outer pyrolytic carbon layer was not effectively wetted by the encapsulating glass matrix. The aqueous durability of the TRISO particle–glass composites was investigated using PCT and MCC-1 tests combined with geochemical modelling. It was found that durability was dependent on silicate and calcium solution saturation. This study provides significant advancements in the preparation of TRISO particle encapsulant waste forms. The potential for the use of non-borosilicate sintered glass composites for TRISO particle encapsulation has been confirmed, although further refinements are required.

A statistical approach to determine the effects of nuclear glass components on the short and long term glass alteration

International Nuclear Information System (INIS)

Advocat, Th.; Tovena, I.; Vernaz, E.; Larche, F.; Phan Tan Luu, R.

1997-01-01

The experimentation plan methodology may be a powerful tool to design statistical models able to calculate quantitative leach rates as a function of glass compositions. The investigation discussed here implemented an experimentation plan methodology covering a wide range of glass composition variations with two major objectives: - identify the effects of the main component oxides of R7T7 glass (SiO 2 , B 2 O 3 , Al 2 O 3 , Na 2 O+Li 2 O, additive oxides, Fission Products oxides, actinide oxides) on the dissolution rates far and close to equilibrium; - and develop a statistical model relating the measured initial dissolution rate with the selected oxide compositions. A second-degree silica-based model was developed to express the initial dissolution rates at 100 deg C according to the oxide weight percentage of 6 major oxides or oxide groups. The model is qualified by comparison with independent experimental results. For the short term, far from saturation conditions (confined media), plotting the effects of the oxides clearly showed that SiO 2 , Al 2 O 3 and the additive oxides enhanced glass durability at 100 deg C, while B 2 O 3 and Na 2 O+Li 2 O diminished the initial corrosion resistance. For the long term, plotting the effects of the oxides showed that additive oxides, Al 2 O 3 and SiO 2 enhanced glass durability at 90 deg C, in that order, while Na 2 O+Li 2 O still diminished the corrosion resistance. The fission products and actinide oxides and ZrO 2 have apparently a slight beneficial effect on the glass durability, under saturation conditions (long term). (author)

From glass structure to its chemical durability

International Nuclear Information System (INIS)

Angeli, F.

2009-01-01

The author gives an overview of his research activities. He more precisely reports studies related to glass structure based on nuclei observed by NMR and present in glasses of interest for nuclear activities. He discusses the influence of chemical composition on structure, and discusses information which can be extracted from network formers (Al, B) and modifiers (Na, Ca), and from oxygen present in the network linkages of oxide glasses. He discusses the different experimental and modelling approaches which enable structural and morphological information to be obtained at a mesoscopic scale. The last part deals with the investigation of the long term behaviour of confinement matrices (glassy matrix for medium-activity wastes, ceramic matrix)

New functionality of chalcogenide glasses for radiation sensing of nuclear wastes

International Nuclear Information System (INIS)

Ailavajhala, M.S.; Gonzalez-Velo, Y.; Poweleit, C.D.; Barnaby, H.J.; Kozicki, M.N.; Butt, D.P.; Mitkova, M.

2014-01-01

Highlights: • Study of thin film chalcogenide glasses under gamma radiation and a proposed radiation sensor design. • Structural changes were observed at various radiation doses. • Formation of Ag 2 Se in Se depleted glasses with sufficient radiation dose. • In conventional semiconductor chip environment, the proposed sensor has a linear current vs. dose behavior up to 600 J/cm 2 . - Abstract: Data about gamma radiation induced effects in Ge 40 Se 60 chalcogenide thin films and radiation induced silver diffusion within these are presented. Blanket films and devices were created to study the structural changes, diffusion products, and device performance. Raman spectroscopy, X-ray diffraction, current vs. voltage (I–V) and impedance measurements expound the behavior of Ge 40 Se 60 glass and silver diffusion within this glass under radiation. Raman study shows that there is a decrease in the area ratio between edge shared and corner shared structural units revealing structural reorganization occurring in the glasses as a result of gamma radiation. X-ray diffraction studies revealed that with sufficiently radiation dose it is also possible to create Ag 2 Se in selenium-depleted systems. Oxidation of the Ge enriched chalcogenide backbone is confirmed through the electrical performance of the sensing elements based on these films. Combination of these structural and diffusion products influences the device performance. The I–V behavior is characterized by increase in current and then stabilization as a function of radiation dose. Additionally, device modeling is also presented using Silvaco software and analytical methods to shed light on the device behavior. This type of sensor design and material characterizations facilitate in improving the radiation sensing capabilities of silver containing chalcogenide glass thin films

New functionality of chalcogenide glasses for radiation sensing of nuclear wastes

Energy Technology Data Exchange (ETDEWEB)

Ailavajhala, M.S., E-mail: m.ailavajhala@gmail.com [Department of Electrical and Computer Engineering, Boise State University, Boise, ID 83725 (United States); Gonzalez-Velo, Y. [School of Electrical, Computer and Energy Engineering, Arizona State University, Tempe, AZ 85287-5706 (United States); Poweleit, C.D. [Department of Physics, Arizona State University, Tempe, AZ 85287-5706 (United States); Barnaby, H.J.; Kozicki, M.N. [School of Electrical, Computer and Energy Engineering, Arizona State University, Tempe, AZ 85287-5706 (United States); Butt, D.P. [Department of Materials Science and Engineering, Boise State University, Boise, ID 83725 (United States); Mitkova, M., E-mail: maheshailavajhala@u.boisestate.edu [Department of Electrical and Computer Engineering, Boise State University, Boise, ID 83725 (United States)

2014-03-01

Highlights: • Study of thin film chalcogenide glasses under gamma radiation and a proposed radiation sensor design. • Structural changes were observed at various radiation doses. • Formation of Ag{sub 2}Se in Se depleted glasses with sufficient radiation dose. • In conventional semiconductor chip environment, the proposed sensor has a linear current vs. dose behavior up to 600 J/cm{sup 2}. - Abstract: Data about gamma radiation induced effects in Ge{sub 40}Se{sub 60} chalcogenide thin films and radiation induced silver diffusion within these are presented. Blanket films and devices were created to study the structural changes, diffusion products, and device performance. Raman spectroscopy, X-ray diffraction, current vs. voltage (I–V) and impedance measurements expound the behavior of Ge{sub 40}Se{sub 60} glass and silver diffusion within this glass under radiation. Raman study shows that there is a decrease in the area ratio between edge shared and corner shared structural units revealing structural reorganization occurring in the glasses as a result of gamma radiation. X-ray diffraction studies revealed that with sufficiently radiation dose it is also possible to create Ag{sub 2}Se in selenium-depleted systems. Oxidation of the Ge enriched chalcogenide backbone is confirmed through the electrical performance of the sensing elements based on these films. Combination of these structural and diffusion products influences the device performance. The I–V behavior is characterized by increase in current and then stabilization as a function of radiation dose. Additionally, device modeling is also presented using Silvaco software and analytical methods to shed light on the device behavior. This type of sensor design and material characterizations facilitate in improving the radiation sensing capabilities of silver containing chalcogenide glass thin films.

Platinoids and molybdenum in nuclear waste containment glasses: a structural study; Les platinoides et le molybdene dans des verres d'interet nucleaires: etude structurale

Energy Technology Data Exchange (ETDEWEB)

Le Grand, M. [CEA/VALRHO - site de Marcoule, Dept. de Recherche en Retraitement et en Vitrification (DRRV), 30 - Marcoule (France)]|[Paris-7 Univ., 75 (France)

2000-07-01

This work deals with the structure of borosilicate nuclear glasses and with some relationships between structure and macroscopic properties. Two types of elements which may disturb the industrial process - platinoids (Ru and Pd) and molybdenum - are central to this work. Platinoids induce weak modifications on the structure of the glass, causing a depolymerization of the glassy network, an increase of the {sup [3]}B/{sup [4]}B ratio and a modification of the medium range order around Si between 3.3 and 4.5 angstrom. The modifications of viscosity and density induced by platinoids in the glass are not due to the structural effect of the platinoids. The increase of viscosity is attributed to needle shaped RuO{sub 2}. It can be moderated by imposing reducing conditions during the elaboration of the glass. The slight difference between experimental and calculated densities is due to the increase of the volume percentage of bubbles in the glass with increasing platinoid content. Mo is either present in the glass as molybdic groupings, or mobilized in chemically complex molybdic crystalline phases. The chemical composition and mineralogy of these phases has been obtained using electronic microprobe data and XRD with Rietveld analysis. The distribution of the different elements between the crystalline phases and the glass is strongly influenced by the structural role of the various cations in the glass. The Mo present in the glass appears as MoO{sub 4} tetrahedra, independent of the borosilicate network. The formation of the crystalline phases can be explained by the existence of a precursor in which the MoO{sub 4} tetrahedra are concentrated in rich alkali and earth-alkali bearing areas of the glass. (author)

Understanding the effects of alpha self-irradiation on the glass structure by coupling spectroscopic studies and atomistic modeling

International Nuclear Information System (INIS)

Bureau, G.

2008-12-01

Our objective was to assess irradiation effects on nuclear containment glass in order to guarantee glass performance when subjected to high alpha decay doses. Experimental studies and molecular dynamics modeling provided a better understanding of the impact of cumulative alpha decay on the structural behavior of complex nuclear glass formulations and of simplified glass models. A mechanism typical of sodium borosilicate glass was identified in response to nuclear interactions or ballistic collisions. The glass local order is slightly modified by the conversion of a fraction of the boron atoms from coordination number IV to III, releasing charge-compensating alkali ions that become available as network modifiers, and resulting in a slight increase in the number of non bridging oxygen atoms. The medium-range order shifts toward increasing disorder in the glass as indicated by broadening of the angular, radial, and size distributions. A model of accumulated quasi-thermal quenching is proposed to account for these changes, based on the two steps describing the reaction of the glass to the alpha decay recoil nucleus: a cascade generates a ballistic phase that completely destabilizes the glass structure with no short and medium-range order, resulting in the loss of the initial structure; glass reconstruction is controlled only by the 'quenching rate' in the displacement cascade, i.e. by its thermal history and the corresponding relaxation options. From this standpoint the final glass structure is the consequence of the ballistic changes and the regenerative capacity of the glass structure, resulting in a higher fictive-temperature glass corresponding to the structural changes identified in this study. (author)

The leaching of radioactivity from highly radioactive glass blocks buried below the water table: fifteen years of results

International Nuclear Information System (INIS)

Merritt, W.F.

1976-03-01

The results from two test burials of high-level fission products incorporated into nepheline syenite glass indicate that the nuclear wastes from fuel processing for a 30,000 MWe nuclear power industry could be incorporated into such glass and stored beneath the water table in the waste management area of Chalk River Nuclear Laboratories (CRNL) without harm to the environment. (author)

Influence of processing conditions on the glass-crystal transition into borosilicate glasses