Gamma-ray shielding and structural properties of PbO-SiO2 glasses

Gamma-ray attenuation coefficients have been determined experimentally using a narrow beam transmission method for the xPbO(1−x)SiO₂ (x = 0.45-0.70) glass system at 662, 1173 and 1332 keV photon energies. These values have also been obtained theoretically using the ‘mixture rule’ and the ‘XCOM’ computer software. The results have been used to calculate half value layer parameters. Gamma-ray shielding properties of PbO-SiO₂ glass samples have been compared with standard radiation shielding concretes. The molar volume, FTIR and acoustic investigations have been used to study the structural properties of the prepared glass system.     

Physical properties of an alumino-silicate waste form for cesium and strontium

Nuclear fuel reprocessing will be required to sustain nuclear power as a baseload energy supplier for the world. New reprocessing schemes offer an opportunity to develop a better strategy for recycling elements in the fuel and preparing stable waste forms. Advanced strategies could create a waste stream of cesium, strontium, rubidium, and barium. Some physical properties of a waste form containing these elements sintered into bentonite clay were evaluated. We prepared samples loaded to 27% by mass to a density of approximately 3 g/cm³. Sintering temperatures of up to 1000 °C did not result in volatility of cesium. Instead, the crystallinity noticeably increased in the waste form as temperatures increased from 600 to 1000 °C. Assemblages of silicates were formed. Significant water evolved at approximately 600 °C but no other gases were generated at higher temperatures.     

Sol-gel silica coatings for the protection of cultural heritage glass

To slow down the phenomenon of weathering of ancient stained glass, the present work proposes the coating of glass. Several recipes have been tested in the past. The coatings are made of sol-gel silica prepared with different catalysts like H⁺, Pb(II), Sn(IV) and without catalyst. All the investigated samples show a good adhesion of the coating to the glasses used to simulate the behaviour of ancient artefacts and a good resistance to ageing tests.In the present work, some tests on Au and Ti surfaces coated with sol-gel silica were done, in order to investigate the adhesion of a H⁺ sol-gel coating to the surface of metals. In fact, in the case of glassy mosaics called “golden leaf tessera”, the adhesion of the glass to the metal is a critical point, due to the fact that the weathering of such tesserae causes the detaching of the “cartellina” (small glass layer) from the metal leaf. The main techniques adopted to investigate are optical microscopy (OM), X-ray photoelectron spectroscopy (XPS), atomic force microscopy (AFM) and Rutherford backscattering spectroscopy (RBS).     

Evaluation of alloy 690 process pot at the contact with borosilicate melt pool during vitrification of high-level nuclear waste

Understanding the material behaviour under service conditions is essential to enhance the life span of alloy 690 process pot used in vitrification of high-level nuclear waste. During vitrification process, interaction of alloy 690 with borosilicate melt takes place for substantial time period. Present experimental studies show that such interactions may result in Cr carbide precipitation along grain boundaries, Cr depletion in austenitic matrix and intergranular attack close to alloy 690/borosilicate melt pool interfaces. Widths of Cr depleted zone within alloy 690 is found to follow kinetics of the type x = 10.9 × 10⁻⁶ + 1 × 10⁻⁸t^1/2 m. Based on the experimental results it is recommended that compositional modification of alloy 690 process pot adjacent to borosilicate melt pool need to be considered seriously for any efforts towards reduction and/or prevention of process pot failures.     

Development of oxides dispersion strengthened steels for high temperature nuclear reactor applications

By introducing a dispersion of nanosized yttrium oxides particles into a steel matrix, the upper temperature limit in mechanical creep strength can be enhanced in temperature by 100 K at least. Production routes for the production of a new class of oxides dispersion strengthened (ODS) steels are investigated within this work. Preliminary results obtained when doping pure iron matrix phase with two types of yttrium oxides (Y₂O₃) nanoparticles (commercial as well as laboratory fabricated nanopowder) are presented. The twofold purpose of this work is firstly to obtain a comparative analysis between the commercial and the laboratory fabricated Y₂O₃ nanopowder used to produce the doped iron, and secondly to demonstrate the feasibility of new production route by observing the nanostructure of the first test batches with pure iron. Observations are carried out with transmission electron microscopy (TEM) to determine the size distribution of the particles in the powder, while glow discharge optical emission spectroscopy (GDOES) and high resolution-scanning electron microscopy (HR-SEM) are used to analyze the chemical composition and the homogeneity of the produced doped iron. It is demonstrated, that even with small size particles nanopowder fabricated in the laboratory, the distribution is fairly homogeneous compared to the one obtained with a relatively large particles commercial nanopowder, confirming the feasibility of the new production route.     

Effect of additive ions on the optical density and stability of the color centers induced by X-ray irradiation in soda-lime silicate glass

To apply radiation-induced coloration of glasses as a reversible glass-coloring technique, we studied the influence of various additive ions incorporated into a soda-lime silicate glass on the optical density and stability of the color induced by X-ray irradiation. Absorption spectra before and after irradiation are discussed in the comparison with those of the undoped soda-lime silicate glass. Additive ions were incorporated as metal oxides, namely TiO₂, V₂O₅, Fe₂O₃, ZnO, Ga₂O₃, GeO₂, ZrO₂, Nb₂O₃, MoO₃, Ag₂O, In₂O₃, SnO, SnO₂, CeO₂, Eu₂O₃, Ta₂O₅, WO₃ and Bi₂O₃. Among them, TiO₂, GeO₂, Nb₂O₃, MoO₃, Ag₂O, In₂O₃, Eu₂O₃, Ta₂O₅, WO₃ and Bi₂O₃ have a large effect on optical density. The optical densities in the visible region for glasses doped with these oxides were much stronger than for undoped soda-lime silicate glass. On the other hand, incorporation of Fe₂O₃, SnO and CeO₂ reduced the optical density. Over longer periods the coloration of the undoped glass was one of the most stable while those of the Fe₂O₃, SnO and CeO₂-doped glasses soon faded.     

Synthesis of dysprosium and cerium nitrides by a mechanically induced gas-solid reaction

A high energy ball milling process was used to produce dysprosium nitride and cerium nitride powders at room temperature. Dysprosium and cerium metal flakes were milled in a 275 kPa nitrogen atmosphere for 24 h at ambient temperatures. X-ray diffraction confirmed the formation of phase pure dysprosium nitride and cerium nitride powders. The median particle size of the resultant dysprosium nitride was measured as 4 μm using a laser scattering technique. The particle size of the cerium nitride was not measured due to its reactive nature.     

M2N nitride phases of 9% chromium steels for nuclear applications

M₂N nitride phases of 9% chromium steels with an extra-low carbon content have been investigated using a transmission electron microscope and an energy-dispersive X-ray (EDX) spectroscopy. The steel samples were normalized for 1 h at 1050 °C and then tempered at 600-780 °C for 30 min to 5 h followed by an air cooling. Through the analyses of the electron micro-diffraction patterns and EDX data for the precipitate particles on the extracted carbon replica, two types of Cr-rich M₂N nitride phases with the same hexagonal structure but totally different lattice parameters, a = 2.80 Å/c = 4.45 Å and a = 7.76 Å/c = 4.438 Å, were determined in the steels. Four types of Cr-rich M₂N phases with different lattice parameters probably existed in the steels. The M₂N phase revealed a decrease in its Cr content, an increase in its V content as the tempering temperature was increased, and no obvious change in its content for the metal fraction with an increasing tempering time.     

Photon interaction and energy absorption in glass: A transparent gamma ray shield

The effective atomic number, Z_eff, the effective electron density, N_e,eff, and the energy dependence, ED, have been calculated at photon energies from 1 keV to 1 GeV for CaO-SrO-B₂O₃, PbO-B₂O₃, Bi₂O₃-B₂O₃, and PbO-Bi₂O₃-B₂O₃ glasses with potential applications as gamma ray shielding materials. For medium-Z glasses, Z_eff is about constant and equal to the mean atomic number in a wide energy range, typically 0.3 < E < 4 MeV, where Compton scattering is the main photon interaction process. In contrast, for high-Z glasses there is no energy region where Compton scattering is truly dominating. Heavy-metal oxide glasses containing PbO and/or Bi₂O₃ are promising gamma ray shielding materials due to their high effective atomic number and strong absorption of gamma rays. They compare well with concrete and other standard shielding materials and have the additional advantage of being transparent to visible light. The single-valued effective atomic number calculated by XMuDat is approximately valid at low energies where photoelectric absorption is dominating.     

Micro-PIXE-RBS methods highlighting the influence of phosphorus on the in vitro bioactivity of sol-gel derived glass particles in the SiO2-CaO-P2O5 system

Ion beam analysis methods were used to characterize the interface of bioactive glasses with surrounding biological fluids. Glass particles in the SiO₂-CaO and SiO₂-CaO-P₂O₅ compositions were made by sol-gel processing and soaked in biological fluids for periods up to 4 days. The surface changes were characterized using PIXE-RBS, which are efficient methods for multielemental analysis and accurate trace elements quantification. Elemental maps of major and trace elements were obtained at a micrometer scale and revealed the bone bonding ability of the materials. Glass particles are quickly coated with a thin calcium phosphate-rich layer containing traces of magnesium. After a few days, SiO₂-CaO-P₂O₅ glass particles are entirely changed into calcium phosphates, whereas SiO₂-CaO particles exhibit a different behavior: the previously Ca-P enriched periphery has been dissolved and glass particles consist of a silicate network. Calculation of the Ca-P atomic ratios at the glass/biological fluids interface provides us with an explanation for this: an enduring apatitic phase seems to be formed at the periphery of SiO₂-CaO-P₂O₅ glass particles. Presence of phosphorus in the glass matrix thus has an influence on the amplitude and the kinetics of reaction of the bioactivity process. It might result in an improved chemical bond with living tissues.     

Effects of 80 KeV proton radiation on the optical properties and microstructure of type-GG17 glass as rubidium lamp envelope

Effect of 80 KeV proton radiation with fluences of 10¹⁴-2 × 10¹⁶ protons/cm² on the optical properties and microstructure of type-GG17 borosilicate glass as a candidate material for rubidium spectra lamp envelope was investigated. The change in microstructure before and after proton radiation was evaluated by means of UV-vis-NIR spectroscopy, X-ray photoelectron spectroscopy (XPS) and positron annihilation lifetime spectrometer (PALS). The experimental results show that under radiation of 80 KeV some broken Si-O bonds were induced and thus the number of non-bridge oxygen (NBO) increased. These irradiation-induced defects caused the degeneration of optical property and loosened glass structure which accelerates rubidium diffusion into lamp envelop.     

TEM investigations of MN nitride phases in a 9% chromium ferritic/martensitic steel with normalization conditions for nuclear reactors

The investigations on the precipitate phases in a 9%Cr ferritic/martensitic steel under different normalization conditions have been made by using a transmission electron microscope and an energy-dispersive X-ray spectroscopy. Hot-rolled steel samples were normalized at 1050-1200 °C for 1-2 h followed by an air cooling to room temperature. MN vanadium nitride precipitates with a plate-like morphology and a chemical formula of about (V_0.4Nb_0.4Cr_0.2)N have been observed at triple junctions, grain boundaries and within matrix in the steel samples normalized at 1050-1150 °C for 1-2 h, but they were dissolved out at 1200 °C within 1 h. Vanadium nitride is a stable phase at 1050 °C according to thermocalc prediction of equilibrium phases in the steel. With increasing normalizing temperature and time, there was no a striking change in the chemical composition of metallic elements in the MN phase, but a considerable increase in the size of the MN precipitate.     

Determination of bulk etch rate for CR-39 nuclear track detectors using an X-ray fluorescence method

The thickness of a CR-39 detector is determined using an energy dispersive X-ray fluorescence (EDXRF) method of analysis. The method is based on exciting a suitable target and measuring the intensity of its fluorescence X-ray lines passing through the CR-39 sample in a fixed geometry. By properly selecting the target material, the method succeeds in assessing the thickness change of CR-39 detectors etched for different time intervals. The bulk etch rate (V_b) may thus be obtained, which is an important parameter for any solid state nuclear track detector. Application of the EDXRF method yielded a value of V_b = (2.01 ± 0.04) μm h⁻¹ for etching in a 6 N NaOH solution at 75 °C. This value agrees with the bulk etch rate of (1.90 ± 0.03) μm h⁻¹, obtained by the conventional mass-change method.     

Investigation on boron carbide oxidation for nuclear reactor safety: Experiments in highly oxidising conditions

The oxidation kinetics of boron carbide pellets were investigated in steam/argon mixtures in the temperature range 1200-1800 °C for steam partial pressures between 0.2 and 0.8 bar and total flows (steam + argon) between 2.5 and 10 g/min resulting in gas velocities from 1.01 to 5.34 m/s. A kinetic model for boron carbide pellet oxidation depending on temperature, steam partial pressure and flow velocity is obtained. The activation energy of the oxidation process was determined to be 163 ± 8 kJ/mol. The strong influence of temperature and steam partial pressure on the boron carbide oxidation kinetics is confirmed. The obtained data suggest the coexistence of two kinetic regimes, one at 1200 °C and the other at 1400-1800 °C, with different dependence on steam partial pressure.     

Literature review of thermal and radiation performance parameters for high-temperature, uranium dioxide fueled cermet materials

High-temperature fissile-fueled cermet literature was reviewed. Data are presented primarily for the W-UO₂ as this was the system most frequently studied; other reviewed systems include cermets with Mo, Re, or alloys as a matrix. Failure mechanisms for the cermets are typically degradation of mechanical integrity and loss of fuel. Mechanical failure can occur through stresses produced from dissimilar expansion coefficients, voids created from diffusion of dissimilar materials or formation of metal hydride and subsequent volume expansion. Fuel loss failure can occur by high temperature surface vaporization or by vaporization after loss of mechanical integrity. Techniques found to aid in retaining fuel include the use of coatings around UO₂ fuel particles, use of oxide stabilizers in the UO₂, minimizing grain sizes in the metal matrix, minimizing impurities, controlling the cermet sintering atmosphere, and cladding around the cermet.     

Effect of sintering conditions on the microstructure and mechanical properties of ZrN as a surrogate for actinide nitride fuels

Pellets of sintered ZrN were studied to optimize the mechanical properties and microstructures needed in nitride fuel pellets, using ZrN as a surrogate for actinide nitrides and as potential component in low fertile and inert matrix fuels. Samples were prepared via sintering in either Ar or N₂ (with and without 6% H₂) and at 1300 °C or 1600 °C. A significant difference in the hardness was measured ranging from 1000 (Kg/mm²) in samples sintered at 1600 °C in argon to 100 (Kg/mm²) in samples sintered at 1300 °C in nitrogen. Samples with 6% hydrogen added to the sintering environment experienced a decrease in hardness, as well as an increase in intergranular cracking as compared to samples sintered without hydrogen, suggesting hydrogen embrittlement. Grain size was more uniform in samples sintered in pure Ar as compared to Ar-H₂, while the latter had a larger fraction of high angle grain boundaries than the former. Cracking around indents had a clear tendency to follow high angle boundaries, which were found to be intrinsically weak in ZrN.     

Aging effects on microstructural and mechanical properties of select refractory metal alloys for space-reactor applications

Refractory alloys based on niobium, tantalum and molybdenum are potential candidate materials for structural applications in proposed space nuclear reactors. Long-term microstructural stability is a requirement of these materials for their use in this type of creep dominated application. Early work on refractory metal alloys has shown aging embrittlement occurring for some niobium and tantalum-base alloys at temperatures near 40% of their melting temperatures in either the base metal or in weldments. Other work has suggested microstructural instabilities during long-term creep testing, leading to decreased creep performance. This paper examines the effect of aging 1100 h at 1098, 1248 and 1398 K on the microstructural and mechanical properties of two niobium (Nb-1Zr and FS-85), tantalum (T-111 and ASTAR-811C) and molybdenum (Mo-41Re and Mo-47.5Re) base alloys. Changes in material properties are examined through mechanical tensile testing coupled with electrical resistivity changes and microstructural examination through optical and electron microscopy analysis.     

Multiple lines of evidence for performance of the canister and waste form in long-term nuclear waste disposal: Reviews

The time scales required for nuclear waste disposal are very large compared with those for other engineering endeavors. Because of this, there are many uncertainties associated with the quantitative performance assessment of canisters containing high-level radioactive waste in a waste form. Multiple lines of evidence can be helpful in building confidence in the long-term behavior (corrosion and dissolution) of the canister and waste form. These lines of evidence are derived from long-term supports and probabilistic models and developed based on shorter term tests, bounding and conservative approaches, and available observations on natural analogs. This paper presents the progress made for important lines of evidence considered in quantitatively assessing radionuclide release behavior from canisters and waste forms. This paper considers risk-significant issues for canisters and waste forms (i.e., risk informed approach) in the probabilistic performance assessment of the disposal system which has also other components such as geology and hydrology.     

The design and calibration of a phantom for depth profiling measurements of entrained radioactivity in silica-based media

The development of a phantom which replicates the effect of concrete on radioactivity entrained within it is described. The phantom was designed as a basis on which methods can be developed to measure the depth of radioactive contamination in the concrete of defunct nuclear facilities. In particular, this apparatus has been used to validate a differential attenuation method for the profiling of radioactive contamination at depth. Entrained radioactive contamination is a significant issue in defunct nuclear facilities where in situ, non-destructive assay of radioactive waste arisings is a routine requirement.The phantom comprises a polymethylmethacrylate structure filled with high-purity silica-sand which, for the purposes of the application, is an effective analogue of fully-hydrated concrete paste. A void was created within the silica sand which incorporates a sliding mechanism for the insertion of a radioactive source to a required depth. The sealed source represents the entrained radioactivity in the phantom but is also specifiable, removable and poses no long-term contamination risk beyond the expected life of the apparatus. The remainder of the void either side of the source is filled with silica-sand to complete the homogeneity of the phantom. The void was situated near the front of the phantom constructed at a 5.14° angle with respect to the front scanning surface; thus the apparent depth within the silica-sand can be varied by changing the position of the source along the phantom's void.The steps taken to develop the concrete phantom are described. The design has been validated with a set of radiation transport simulations affording comparison with an exemplar of the concrete found used in nuclear facilities. Some initial results from measurements taken with the phantom and a caesium-137 γ-radiation source in combination with a sodium iodide radiation detector are provided. These measurements are used to validate the differential attenuation method and compared with data from previous measurement attempts with concrete slices.