Penetration of tritiated water vapor through hydrophobic paints for concrete materials

Behavior of tritium transfer through hydrophobic paints of epoxy and acrylic-silicon resin was investigated experimentally. The amounts of tritium permeating through their paint membranes were measured under the HTO concentration condition of 2–96 Bq/cm³. Most of tritium permeated through the paints as a molecular form of HTO at room temperature. The rate of tritium permeating through the acrylic-silicon paint was correlated in terms of a linear sorption/release model, and that through the epoxy paint was controlled by a diffusion model. Although effective diffusivity estimated by a diffusion model was obtained 1.1 × 10⁻¹³–1.8 × 10⁻¹³ m²/s for epoxy membranes at the temperature of 21–26 °C, its value was found to be hundreds times larger than that for cement-paste coated with epoxy paint. Hence, resistance of tritium diffusion through interface between cement-paste and the epoxy paint was considered to be the most effective in the overall tritium transfer process. Clarification of tritium transfer behavior at the interface is important to understand the mechanism of tritium transfer in concrete walls coated with various paints.     

Effect of hydrophobic paints coating for tritium reduction in concrete materials

The effects of hydrophobic paint coating on a concrete material of cement paste on the tritium transport are investigated. The cement paste is coated with two kinds of paints, acrylic-silicon resin paint and epoxy resin paint. We investigated the amount of tritium trapped in the samples exposed to tritiated water vapor by means of sorption and release. It was found that both the hydrophobic paints could reduce effectively tritium permeation during 50 days exposure of tritiated water vapor. The effect of tritium reduction of the epoxy paint was higher than that of silicon while the amount of tritium trapped in the epoxy paint was larger than that of silicon due to difference of the structure. Based on an analysis of a diffusion model, the rate-determining step of tritium migration through cement paste coated with the paints is diffusion through the paints respectively. It was found that tritium was easy to penetrate through silicon because there were many pores or voids in the silicon comparatively. In the case of tritium released from the epoxy paint, it is considered that tritium diffusion in epoxy is slow due to retardation by isotope exchange reaction to water included in epoxy paint.     

Penetration behavior of water solution containing radioactive species into dried concrete/mortar and epoxy resin materials

Penetration behavior of radionuclides such as ¹³⁷Cs into dried concrete material, dried mortar material and epoxy paint for a few dozen days was observed using a solution containing fission products extracted from irradiated fuels to obtain fundamental information on the radionuclide penetration rate and depth. Hardly any radionuclides could penetrate into the epoxy paint. The radionuclide solution penetrated into concrete and mortar materials to a depth of a few millimeters for a few dozen days. The penetration behavior observed near the surface of concrete and mortar materials was similar to the diffusion of nuclides in media such as water-saturated concrete, bentonite and cement materials.     

Experiment and Modeling for Solvent Leaching from Paint Matrix Considering Equilibrium

The leaching behavior of remaining solvent from an epoxy paint coating when it is submerged in water was experimentally studied. A leach kinetics model considering the equilibrium of the solvent concentration in the paint matrix and in water was developed. Three model parameters, equilibrium constant K, leaching rate k_d , and initial concentration of the solvent in the paint C⁰ _p, were evaluated based on the experimental results, and empirical correlation equations for them were obtained. The model showed good qualitative and quantitative agreement with the observed evolution of the leached solvent mass in the present experiment. Also, the model showed consistency with experimental data in a different configuration.     

Measurement of the leaching rate of radionuclide Cs from the solidified radioactive sources in Portland cement mixed with microsilica and barite matrixes

Portland cement was mixed with radionuclide ¹³⁴Cs to produce low-level radioactive sources. These sources were surrounded with cement mixed with different materials like microsilica and barite. The leaching rate of ¹³⁴Cs from the solidified radioactive source in Portland cement alone was found to be 4.481 × 10⁻⁴ g/cm² per day. Mixing this Portland cement with microsilica and with barite reduced significantly the leaching rate to 1.091 × 10⁻⁴ g/cm² per day and 3.153 × 10⁻⁴ g/cm² per day for 1 wt.% mixing, and to 1.401 × 10⁻⁵ g/cm² per day and 1.703 × 10⁻⁴ g/cm² per day for 3 wt.% mixing, respectively. It was also found that the application of a latex paint reduced these leaching rates by about 6.5%, 20.3% and 13.3% for Portland cement, cement mixed with microsilica and with barite, respectively. The leaching data were also analyzed using the polynomial method. The obtained results showed that cement mixed with microsilica and with barite can be effectively used for radioactive sources solidification.     

Characterization of advanced cyanate ester/epoxy insulation systems before and after reactor irradiation

Insulation systems for fusion magnets have to operate in a harsh environment, especially also under intense radiation. Over the past years, cyanate ester (CE) resins have been playing an increasingly important role because of their enhanced temperature and radiation resistance compared to conventional epoxy resins. Blending CE with epoxy resins offers the possibility to manufacture radiation resistant insulations at a low price compared to pure CE materials. Therefore, it is of special interest to study the influence of the CE content and of the epoxy resin on the mechanical properties to find materials, which are suitable and economically reasonable for the specific demands of such magnets.In this study R-glass fiber/Kapton reinforced cyanate ester/epoxy blends with different CE content were investigated. Each material was exposed to conditions matching those expected for the ITER TF coil insulation as closely as possible. In order to characterize the mechanical properties, short-beam shear and static tensile tests were carried out at 77 K prior to and after irradiation to fast neutron fluences of up to 5 × 10²² m⁻² (E > 0.1 MeV), in the TRIGA reactor (Vienna) at ambient temperature (340 K). In addition, tension-tension fatigue measurements were performed in the load-controlled mode to simulate the pulsed operation conditions of ITER.     

Study of electron beam curing process using epoxy resin system

Polymeric matrix composite (PMC) has been used in engineering applications instead of metal in the last few years, due to its corrosion resistance and excellent relation between tensile strength/density and elastic modulus/density. However, PMC materials cured by thermal process require high temperature and are time-consuming. The electron beam (EB) curing technology allows its use at room temperature and reduced curing times, and this is one of the main advantages over thermal technology. The aim of this work is to investigate electron beam curable epoxy formulations to use in filament winding processes to produce composite material with similar or better properties than thermal curable composites. The study has been made with commercial epoxy resins and cationic initiators. The epoxy resin samples were irradiated for few minutes with total dose of 150 kGy. The glass transition temperatures (T_g) were determined by dynamic mechanical analyzer (DMA) and the result was 137 °C. The thermal process was carried out in a furnace following three steps: 4 h at 90 °C, increasing temperature from 90 °C to 130 °C during 4 h and 12 h at 130 °C. The total process time was 20 h. The T_g of this sample was 102 °C.     

Oxygen glow discharge experiment to remove deposited layers and to release trapped hydrogen isotopes in HT-7 superconducting tokamak

An experiment to remove re-deposited layers and to release hydrogen using a glow discharge in oxygen (O-GDC) has been performed in the HT-7 superconducting tokamak. In the absence of magnetic fields, the O-GDC wall conditioning had produced rapid, controlled co-deposit removal. Average removal rates, 5.2 × 10²² H-atoms/h, 5.65 × 10²¹ D-atoms/h and 5.53 × 10²² C-atoms/h, respectively, were obtained during 145 min O-GDC experiment in the pressure range 0.5-1.5 Pa. The corresponding removal rate of co-deposited films was ∼1.19 μm/day (26.5 g/day for carbon) based on an area of 12 m². Compared to thermo-oxidation and O-ICR experiment, high pressure O-GDC wall conditioning promoted the oxidation and improved the C and D atoms removal. In the O-GDC experiment, the removal rates of H-atoms and D-atoms as H₂O, HDO and D₂O were higher than that of H₂ and D₂ by factors of about 20 and 50, respectively. During the 145 min O-GDC experiment, about 14.5% O-atoms were converted into carbon oxides and hydroxides, and about 5.37 × 10²² O-atoms were adsorbed on the walls corresponding to a coverage of 4.5 × 10²¹ O/m² on an wall area of 12 m². In a 100 min helium glow discharge (He-GDC) following the O-GDC experiment, 1.53 × 10²² O-atoms, about 28.5% oxygen retained on the walls, were removed. The removal rate of H-atoms in He-GDC cleaning after O-GDC experiment was lower than that in He-GDC cleaning before O-GDC experiment, which indicates that the O-GDC wall conditioning had effectively reduced hydrogen retention on the walls.     

An experimental study of the shielding characteristics of the dwelling house building materials against gamma radiations in the Central Region of Syria

The shielding properties of the concrete and blocks used for the construction of dwelling houses in the Central Region of Syria (CRS) were measured and studied. The concrete used for the ceiling construction was found to have optimum shielding properties with 0.182 cm⁻¹ (or equivalently 0.0859 cm² g⁻¹) for the linear (mass) attenuation coefficient [L(M)AC]. In addition gamma radiation is attenuated by 73.221% on average, while the blocks used for the walls have smaller LACs (0.082 cm⁻¹ for the bare blocks, and 0.118 cm⁻¹ for the coated ones). Although the LACs for the blocks are smaller than those for the concrete their shielding properties are good to protect from the gamma radiations coming from radioactive or nuclear accidents (78.630% attenuation), even Chernobyl – like disasters, because of their big width (10–12 cm). The LACs were measured by an ionization chamber and simple theoretical calculations have been made to predict the concrete LACs. The calculations showed an average LAC for the six samples equal to 0.1664 cm⁻¹ with 8.47% error with respect to the experimental values.     

Dense plasma focus ion-based titanium nitride coating on titanium

We report the synthesis of titanium nitride coating on a titanium substrate by utilizing energetic nitrogen ions emitted from a 2.3 kJ dense plasma focus device for 30 focus shots. The number of nitrogen ions transferred to the sample by a single ion pulse of about 140 ns duration in the energy interval (40-600 keV) is about 1.09 × 10¹² with a mean energy per ion of 58 keV. The corresponding energy flux delivered to the titanium surface is estimated to be 6.17 × 10¹⁴ keV cm⁻³ ns⁻¹ leading to a transient temperature rise of the top layer of about 5400 K which helps layer growth. The coating is investigated on the basis of its morphological, compositional and hardness properties. X-ray diffraction analysis of the sample reveals the formation of a nanocrystalline titanium nitride coating having (1 1 1) and (2 0 0) plane reflections with an average crystallite size of 40 nm. The compressive residual stresses in the nitride coating have been evaluated to be 2.80 GPa and 6.81 GPa corresponding to (1 1 1) and (2 0 0) plane orientations. A complete restructuring of the manually polished titanium substrate has been observed with appearance of nano-sized multidimensional granular surface morphologies. The thickness of the nitride coating is about 1 μm, whereas the coating has a nitrogen content of 35.35 at.% and 13.78 ± 3.57 wt.% and a surface hardness of 8.19 GPa.     

Radiation shielding of concretes containing different lime/silica ratios

The shielding of γ-rays and fast neutrons by concrete has been studied for concretes containing different lime/silica ratios. Calculations were carried out for six different concrete samples. The total mass attenuation coefficients (μ/ρ, cm² g⁻¹) have been computed at photon energies of 1 keV to 100 GeV using the personal computer software package WinXCom. Also the macroscopic effective fast neutron removal cross-sections (Σ_R, cm⁻¹) have been calculated using MERCSF-N program and the removal cross-section database for all required elements. The obtained results showed that the lime/silica ratio of concrete has significant and insignificant effects on μ/ρ and Σ_R values, respectively.     

Neutron dose transmission measurements for several new concrete samples including colemanite

We have prepared four baryte and four concrete samples having respectively 0%, 5%, 10% and 15% colemanite concentrations. Neutron dose transmission measurements have been done by using a source of mono-energetic neutrons (E_eff = 4.5 MeV–²⁴¹Am–Be). It has been shown that when colemanite percentages of the samples increase, neutron dose transmission values for the samples decrease. It is thus possible to enhance the neutron shielding property of baryte and ordinary concrete by adding different proportions of colemanite.     

Calculation of fast neutron removal cross-sections for some compounds and materials

Shielding materials usually contain some well-known compounds or substances. Thus, construction of a database for the effective fast neutrons mass removal cross-sections, Σ_R/ρ (cm² g⁻¹), for these compounds and materials will facilitate the shielding design problem. In this work, building of such database for different compounds and materials which have been widely used in shielding media has been suggested. Furthermore, a short version of this database was given. It includes the calculated values of Σ_R/ρ (cm² g^–1) and Σ_R (cm⁻¹) for some common compounds found in concrete, Portland cement clinker, iron ores and polymers. By using this database, a quick short hand calculation method for rough estimation of macroscopic removal cross-section Σ_R for these materials can be developed. It is interesting to note that considerable removal cross-section values for PVC, PP, PA6, PA6/6, PA6/10, PA11 and PPS polymers were noticed.     

Development of a method for measuring the ortho-positronium quenching rate in low vapor-pressure gases

A new measurement and analysis method for determining the ortho-positronium collisional quenching rate for low vapor-pressure gases is developed. It uses hydrophobic silica aerogels as highly efficient positronium formation media and microchambers. The value of the normalized positronium quenching rate, ¹Z_eff, for CH₃Br measured by this method is 0.70 ± 0.04.     

Behaviour of chromium steels in liquid Pb-55.5Bi with changing oxygen content and temperature

The behaviour of protective oxide layers on P122 steel and its welds and of ODS steel in liquid Pb_44.5Bi_55.5 (LBE) is examined under conditions of changing temperatures and oxygen concentrations. P122 (12Cr) and its welded joints are exposed to LBE at 550 °C for 4000 h with oxygen concentrations of 10⁻⁶ and 10⁻⁸ wt% (p(O₂) = 8.1 × 10⁻²³ bar and 5.2 × 10⁻²⁷ bar) which change every 800 h. It is found that like in case of constant oxygen concentration of 10⁻⁶ wt% a protective spinel layer (Fe(Fe_1−xCr_x)₂O₄) was maintained on P122 and also on its welded joint. Two experiments with exposure times of 4800 h are conducted on ODS steel, both with temperatures changing from 550 to 650 °C and back every 800 h, one experiment with 10⁻⁶ the other with 10⁻⁸ wt% oxygen in LBE. Both experiments show strong local dissolution attack after 4800 h which is in agreement with the behaviour of ODS in LBE at a constant temperature of 650 °C. However, dissolution attack is less in LBE with 10⁻⁸ wt% oxygen (p(O₂) = 3.0 × 10⁻²⁵ bar).     

Measurement of thermal neutron cross-sections and resonance integrals for Hf(n,γ)Hf and Hf(n,γ)Hf reactions at the Pohang neutron facility

The thermal-neutron cross-sections and the resonance integrals for the ¹⁷⁹Hf(n,γ)^180mHf and the ¹⁸⁰Hf(n,γ)¹⁸¹Hf reactions have been measured by the activation method. The high purity Hf and Au metallic foils within and without a Cd shield case were irradiated in a neutron field of the Pohang neutron facility. The gamma-ray spectra from the activated foils were measured with a calibrated p-type high-purity Ge detector.In the experimental procedure, the thermal neutron cross-sections, σ₀, and resonance integrals, I₀, for the ¹⁷⁹Hf(n,γ)^180mHf and the ¹⁸⁰Hf(n,γ)¹⁸¹Hf reactions have been determined relative to the reference values of the ¹⁹⁷Au(n,γ)¹⁹⁸Au reaction, with σ₀ = 98.65 ± 0.09 barn and I₀ = 1550 ± 28 barn. In order to improve the accuracy of the experimental results, the interfering reactions and necessary correction factors were taken into account in the determinations. The obtained thermal neutron cross-sections and resonance integrals were σ₀ = 0.424 ± 0.018 barn and I₀ = 6.35 ± 0.45 barn for the ¹⁷⁹Hf(n,γ)^180mHf reaction, and σ₀ = 12.87 ± 0.52 barn and I₀ = 32.91 ± 2.38 barn for the ¹⁸⁰Hf(n,γ)¹⁸¹Hf reaction. The present results are in good agreement with recent measurements.     

Enthalpy measurements of La2Te3O9 and La2Te4O11

Enthalpy increment measurements on La₂Te₃O₉(s) and La₂Te₄O₁₁(s) were carried out using a Calvet micro-calorimeter. The enthalpy values were analyzed using the non-linear curve fitting method. The dependence of enthalpy increments with temperature was given as: H°(T) − H°(298.15 K) (J mol⁻¹) = 360.70T + 0.00409T² + 133.568 × 10⁵/T − 149 923 (373 ? T (K) ? 936) for La₂Te₃O₉ and H°(T) − H°(298.15 K) (J mol⁻¹) = 331.927T + 0.0549T² + 29.3623 × 10⁵/T − 114 587 (373 ? T (K) ? 936) for La₂Te₄O₁₁.     

Estimation of minimum detectable concentration of chlorine in the blast furnace slag cement concrete

The Prompt Gamma Neutron Activation Analysis technique was used to measure the concentration of chloride in the blast furnace slag (BFS) cement concrete to assess the possibility of reinforcement corrosion. The experimental setup was optimized using Monte Carlo calculations. The BFS concrete specimens containing 0.8-3.5 wt.% chloride were prepared and the concentration of chlorine was evaluated by determining the yield of 6.11, 6.62, 7.41, 7.79 and 8.58 MeV gamma-rays.The Minimum Detectable Concentration (MDC) of chlorine in the BFS cement concrete was estimated. The best value of MDC limit of chlorine in the BFS cement concrete was found to be 0.034 ± 0.011 and 0.038 ± 0.012 wt.% for 6.11 and 6.62 MeV prompt gamma-rays. Within the statistical uncertainty the lower bound of the measured MDC of chlorine in the BFS cement concrete meets the maximum permissible limit of 0.03 wt.% of chloride set by the American Concrete Institute.     

The dissociation of CH and CH2 molecules in He and N2 at beam energies of 80-250 keV and possible implications for radiocarbon mass spectrometry

Isotopic ratios of ¹⁴C at natural levels can be efficiently measured with accelerator mass spectrometry (AMS). In compact AMS systems, ¹³CH and ¹²CH₂ molecular interferences are destroyed in collisions with the stripper gas, a process which can be described by dissociation cross sections. These dissociation cross sections determine the gas areal density required for sufficient attenuation of the interfering molecular beams, and are therefore key parameters in the effort to further reduce the terminal voltage and thus the size of the AMS system. We measured the dissociation cross sections of ¹³CH and ¹²CH₂ in N₂ and He in the energy range of 80-250 keV. In N₂, cross sections were constant for energies above 100 keV with average values per molecule of (8.1 ± 0.4) × 10⁻¹⁶ cm² for ¹³CH and (9.5 ± 0.5) × 10⁻¹⁶ cm² for ¹²CH₂. In He, cross sections were constant over the full measured range of 80-150 keV with average values of (4.2 ± 0.3) × 10^{− 16} cm² and (4.8 ± 0.4) × 10⁻¹⁶ cm², respectively. A considerable reduction of the terminal voltage from the currently used 200 kV while using N₂ for ¹³CH and ¹²CH₂ molecule dissociation is not possible: the required N₂ areal densities of ∼1.4 μg/cm², consequential angular straggling and a decreasing 1+ charge state fraction would reduce the ion beam transmission too much. This is not the case for He: sufficient molecule dissociation can be obtained with gas densities of ∼0.4 μg/cm², for which angular straggling is relatively small. In addition, the 1+ charge state fraction still increases at lower stripping energies. Thus, the usage of He for stripping and molecule dissociation might allow the development of even smaller ¹⁴C-AMS systems than available today.     

The standard molar enthalpy of formation of CdMoO4

The molar enthalpies of solution of CdMoO₄(s), CdO(s), Na₂ MoO₄(s) and NaF(s) in (10 mol HF(aq) + 4.41 mol H₂O₂(aq)) dm⁻³ have been measured using an isoperibol type calorimeter. From these results and other auxiliary data, the standard molar enthalpy of formation of CdMoO₄(s) has been calculated to be Δ_fH°(298.15 K) = −(1034.3 ± 5.7) kJ mol⁻¹. This value of enthalpy of formation of CdMoO₄(s) agrees well with the estimated enthalpy of formation of this compound. There is no other report on the thermodynamic property measurements on this compound.