Deuterium retention in sintered boron carbide exposed to a deuterium plasma

Depth profiles of deuterium up to a depth of 10 μm have been measured using the D(³He,p)⁴He nuclear reaction in a resonance-like technique after exposure of sintered boron carbide, B₄C, at elevated temperatures to a low energy (≈200 eV/D) and high ion flux (≈10²¹ m⁻² s⁻¹) D plasma. The proton yield was measured as a function of incident ³He energy and the D depth profile was obtained by deconvolution of the measured proton yields using the program SIMNRA. D atoms diffuse into the bulk at temperatures above 553 K, and accumulate up to a maximum concentration of about 0.2 at.%. At high fluences (?10²⁴ D/m²), the accumulation in the bulk plays a major role in the D retention. With increasing exposure temperature, the amount of D retained in B₄C increases and exceeds a value of 2 × 10²¹ D/m² at 923 K. The deuterium diffusivity in the sintered boron carbide is estimated to be D = 2.6 × 10⁻⁶exp{−(107 ± 10) kJ mol⁻¹/RT} m² s⁻¹.     

Ion induced segregation in gold nanostructured thin films on silicon

We report a direct observation of segregation of gold atoms to the near surface regime due to 1.5 MeV Au²⁺ ion impact on isolated gold nanostructures deposited on silicon. Irradiation at fluences of 6 × 10¹³, 1 × 10¹⁴ and 5 × 10¹⁴ ions cm⁻² at a high beam flux of 6.3 × 10¹² ions cm⁻² s⁻¹ show a maximum transported distance of gold atoms into the silicon substrate to be 60, 45 and 23 nm, respectively. At a lower fluence (6 × 10¹³ ions cm⁻²) transport has been found to be associated with the formation of gold silicide (Au₅Si₂). At a high fluence value of 5 × 10¹⁴ ions cm⁻², disassociation of gold silicide and out-diffusion lead to the segregation of gold to defect - rich surface and interface regions.     

Existence states of deuterium irradiated into LiAlO2

The existence states of deuterium in LiAlO₂ were analyzed by in situ IR absorption spectroscopy during irradiation with 3 keV at room temperature. Multiple IR absorption peaks that were related to O-D stretching vibrations were observed, mainly at 2650 cm⁻¹ (O-D_α), 2600 cm⁻¹ (O-D_β), and 2500 cm⁻¹ (O-D_γ). The O-D_α was assigned to the surface O-D. The O-D_β and O-D_γ were interpreted as two distinct O-D states for three candidates: O-D of substitutional D⁺ for Li⁺; O-D of substitutional D⁺ for Al³⁺; and O-D of interstitial D⁺. O-D_β was the dominant O-D state for deuterium irradiated into LiAlO₂, and had higher stability than O-D_γ. Heating after ion irradiation led to the desorption of D₂ and an increase in the intensity of O-D_β, which implies that some of the deuterium irradiated into LiAlO₂ exists in non-O-D states, such as D⁻ captured by F centers.     

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.     

Surface mediated isotope exchange reactions between water and gaseous deuterium

Maintaining isotopic purity of hydrogen is one of the major tasks in tritium processing systems. The work with multiple isotopes and isotopomers is accompanied by isotope exchanges which is often accelerated by catalysts e.g. surfaces of various materials. In this work, densities of D₂O, HDO produced via isotope exchange reactions in the mixture of D₂, H₂, D₂O, H₂O, HD and HDO contained in a stainless steel (type SS304) vessel were measured as a function of time (40-36 000 s) and pressures near 3.5 × 10² Pa, using mass spectrometry. The derived rates of change of the isotopomers densities are described accurately by a postulated kinetic model.     

Neutron radiation effects on metal oxide semiconductor (MOS) devices

The main purpose of this study is to provide the knowledge and data on Deuterium-Tritium (D-T) fusion neutron induced damage in MOS devices. Silicon metal oxide semiconductor (MOS) devices are currently the cornerstone of the modern microelectronics industry. However, when a MOS device is exposed to a flux of energetic radiation or particles, the resulting effects from this radiation can cause several degradation of the device performance and of its operating life. The part of MOS structure (metal oxide semiconductor) most sensitive to neutron radiation is the oxide insulating layer (SiO₂). When ionizing radiation passes through the oxide, the energy deposited creates electron-hole pairs. These electron-hole pairs have been seriously hazardous to the performance of these electronic components. The degradation of the current gain of the dual n-channel depletion mode MOS caused by neutron displacement defects, was measured using in situ method during neutron irradiation. The average degradation of the gain of the current is about 35 mA, and the change in channel current gain increased proportionally with neutron fluence. The total fusion neutron displacement damage was found to be 4.8 × 10⁻²¹ dpa per n/cm², while the average fraction of damage in the crystal of silicon was found to be 1.24 × 10⁻¹². All the MOS devices tested were found to be controllable after neutron irradiation and no permanent damage was caused by neutron fluence irradiation below 10¹⁰n/cm². The calculation results shows that (n,α) reaction induced soft-error cross-section about 8.7 × 10⁻¹⁴ cm², and for recoil atoms about 2.9 × 10⁻¹⁵ cm², respectively.     

Corrosion of alkali-borosilicate waste glass K-26 in non-saturated conditions

Experimental data obtained during long term environmental tests of a nuclear waste alkali-borosilicate glass K-26 in an experimental near-surface repository are examined. Average leaching rates of the radionuclides were calculated: the leach rates gradually diminished from 9.4 × 10⁻⁷ g cm⁻² day⁻¹ over the first year to 2.2 × 10⁻⁷ g cm⁻² day⁻¹ over 16 years of tests. Radionuclide losses obey a square root time dependence indicating a diffusion-controlled release mechanism. The main parameters, which control the corrosion of waste glass K-26 in the near-surface repository, are the effective diffusion coefficient of radiocaesium D_Cs and the rate of glass hydrolysis r_h. Analysis of 16 years experimental data gave D_Cs = 4.5 × 10⁻¹² cm² day⁻¹ and r_h = 0.1 μm years⁻¹. Diffusion is predicted to be dominant for 16.4 years after which diffusion and hydrolytic dissolution are expected to be similarly important. This mixed stage is predicted continue for 262 years after which hydrolytic dissolution will be the dominant mechanism.     

Formation of silicon hydride using hyperthermal negative hydrogen ions (H) extracted from an argon-seeded hydrogen sheet plasma source

An E × B probe (a modified Wien filter) is constructed to function both as a mass spectrometer and ion implanter. The device, given the acronym EXBII selects negative hydrogen ions (H⁻) from a premixed 10% argon-seeded hydrogen sheet plasma. With a vacuum background of 1.0 × 10⁻⁶ Torr, H⁻ extraction ensues at a total gas feed of 1.8 mTorr, 0.5 A plasma discharge. The EXBII is positioned 3 cm distance from the sheet core as this is the region densely populated by cold electrons (T_e ∼ 2 eV, N_e ∼ 3.4 × 10¹¹ cm⁻³) best suited for H⁻ formation. The extracted H⁻ ions of flux density ∼0.26 A/m² are segregated, accelerated to hyperthermal range (<100 eV) and subsequently deposited into a palladium-coated 1.1 × 1.1 cm², n-type Si (1 0 0) substrate held at the rear end of the EXBII, placed in lieu of its Faraday cup. The palladium membrane plays the role of a catalyst initiating the reaction between Si atoms and H⁻ ions simultaneously capping the sample from oxidation and other undesirable adsorbents. AFM and FTIR characterization tests confirm the formation of SiH₂. Absorbance peaks between 900-970 cm⁻¹ (bending modes) and 2050-2260 cm⁻¹ (stretching modes) are observed in the FTIR spectra of the processed samples. It is found that varying hydrogen exposure time results in the shifting of wavenumbers which may be interpreted as changes in the frequencies of vibration for SiH₂. These are manifestations of chemical changes accompanying alterations in the force constant of the molecule. The sample with longer exposure time exhibits an additional peak at 2036 cm⁻¹ which are hydrides of nano-crystalline silicon.     

Permeability of hydrogen and deuterium of Hastelloy XR

Permeation of hydrogen isotope through a high-temperature alloy used as heat exchanger and steam reformer pipes is an important problem in the hydrogen production system connected to be a high-temperature engineering test reactor (HTTR). An experiment of hydrogen (H₂) and deuterium (D₂) permeation was performed to obtain permeability of H₂ and D₂ of Hastelloy XR, which is adopted as heat transfer pipe of an intermediate heat exchanger of the HTTR. Permeability of H₂ and D₂ of Hastelloy XR were obtained as follows. The activation energy E₀ and pre-exponential factor F₀ of the permeability of H₂ were E₀=67.2±1.2 kJ mol⁻¹ and F₀=(1.0±0.2)×10⁻⁸ m³(STP) m⁻¹ s⁻¹ Pa^−0.5, respectively, in the pipe temperature ranging from 843 K (570 °C) to 1093 K (820 °C). E₀ and F₀ of the permeability of D₂ were respectively E₀=76.6±0.5 kJ mol⁻¹ and F₀=(2.5±0.3)×10⁻⁸ m³(STP) m⁻¹ s⁻¹ Pa^−0.5 in the pipe temperature ranging from 943 K (670 °C) to 1093 K (820 °C).     

Radiation damage in ZnO ion implanted at 15 K

Commercial O-face (0 0 0 1) ZnO single crystals were implanted with 200 keV Ar ions. The ion fluences applied cover a wide range from 5 × 10¹¹ to 7 × 10¹⁶ cm⁻². The implantation and the subsequent damage analysis by Rutherford backscattering spectrometry (RBS) in channelling geometry were performed in a special target chamber at 15 K without changing the target temperature of the sample. To analyse the measured channelling spectra the computer code DICADA was used to calculate the relative concentration of displaced lattice atoms.Four stages of the damage evolution can be identified. At low ion fluences up to about 2 × 10¹³ cm⁻² the defect concentration increases nearly linearly with rising fluence (stage I). There are strong indications that only point defects are produced, the absolute concentration of which is reasonably given by SRIM calculations using displacement energies of E_d(Zn) = 65 eV and E_d(O) = 50 eV. In a second stage the defect concentration remains almost constant at a value of about 0.02, which can be interpreted by a balance between production and recombination of point defects. For ion fluences around 5 × 10¹⁵ cm⁻² a second significant increase of the defect concentration is observed (stage III). Within stage IV at fluences above 10¹⁶ cm⁻² the defect concentration tends again to saturate at a level of about 0.5 which is well below amorphisation. Within stages III and IV the damage formation is strongly governed by the implanted ions and it is appropriate to conclude that the damage consists of a mixture of point defects and dislocation loops.     

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.     

Corrosion behaviour of steels in lead-bismuth at 823 K

The corrosion behaviour of the martensitic T91 steel and the austenitic AISI 316L steel was analysed. The steels were immersed in stagnant molten Pb-55.2wt%Bi alloy at 823 K for different exposure times (t = 550-2000 h). The corrosion tests were carried out both under Ar and under Ar-5%H₂ mixture. Under the oxidising conditions (P_O2 = 6 × 10⁻³ Pa), the formation of oxide layers was observed which prevent the penetration of the liquid alloy into the matrix, while under the Ar-5%H₂ mixture (P_O2 = 3.2 × 10⁻²³ Pa), two phenomena occurred: a ‘reactive penetration’ at the liquid alloy/steel interface and the competition between oxidation and penetration.     

Ferromagnetic modification of GaN film by Cu ions implantation

The structural and magnetic properties of Cu⁺ ions-implanted GaN films have been reported. Eighty kiloelectron-volt Cu⁺ ions were implanted into n-type GaN film at room temperature with fluences ranging from 1 × 10¹⁶ to 8 × 10¹⁶ cm⁻² and subsequently annealed at 800 °C for 1 h in N₂ ambient. PIXE was employed to determine the Cu-implanted content. The magnetic property was measured by the Quantum Design MPMS SQUID magnetometer. No secondary phases or clusters were detected within the sensitivity of XRD. Raman spectrum measurement showed that the Cu ions incorporated into the crystal lattice positions of GaN through substitution of Ga atoms. Apparent ferromagnetic hysteresis loops measured at 10 K were presented. The experimental result showed that the ferromagnetic signal strongly increased with Cu-implanted fluence from 1 × 10¹⁶ to 8 × 10¹⁶ cm⁻².     

Surface morphology and deuterium retention in tungsten oxide layers exposed to low-energy, high flux D plasma

Surface morphology and deuterium retention in tungsten oxide layers (WO_3−z, z ? 0.25) grown on polycrystalline and recrystallized W substrates have been examined after exposure to a low-energy (38 eV/D), high flux (10²² D/m² s) D plasma to an ion fluence of 10²⁶ D/m² at various temperatures (up to ∼700 K). Characterization methods used were scanning electron microscopy, X-ray diffraction, Rutherford backscattering spectroscopy, and the D(³He,p)⁴He nuclear reaction analysis. During exposure to the D plasma at temperatures of 340-615 K, a partial reduction of the tungsten oxide takes place in the near-surface layer up to 0.3 μm in depth. Even at around room temperature, deuterium atoms diffuse several micrometers into the tungsten oxide. The high D concentration of about 0.1 D/W observed in the first micrometers below the surface at temperatures below 500 K can be related mainly to D atoms chemically bonded to O atoms. As the exposure temperature increases, the D concentration decreases, reaching about 2 × 10⁻⁴ D/W at 615 K. At plasma exposure temperatures of about 700 K, the oxide layer shrinks and loses a large fraction of oxygen.     

Oxygen removal with D2-ICR cleanings after oxidation experiment in HT-7

The effectiveness of D₂-ICR cleaning with a pressure up to 0.18 Pa was surveyed on hot walls of 400-470 K for oxygen removal after oxidation experiment in the HT-7 superconducting tokamak. The oxygen removal rate in D₂-ICR cleaning was about (7-9) × 10²¹ O-atoms/h, about 5-10 times higher that that He-ICR cleanings before and after the D₂-ICR cleaning. In about 130 min He-ICR and D₂-ICR cleanings, about 8 × 10²¹ O-atoms were removed. After the cleanup, a lot of water still retained in HT-7 vessel and 67 disruptive plasmas were required before obtaining normal plasma. In present experiment, the recovered plasmas were easily controlled and much better than previous report in HT-7 oxidation experiments.     

Study of silver diffusion in silicon carbide

Diffusion of silver in 6H-SiC and polycrystalline CVD-SiC was investigated using α-particle channeling spectroscopy and electron microscopy. Fluences of 2 × 10¹⁶ cm⁻² of ¹⁰⁹Ag⁺ were implanted with an energy of 360 keV at room temperature, at 350 °C and 600 °C, producing an atomic density of approximately 2% at the projected range of about 110 nm. The broadening of the implantation profile and the loss of silver through the front surface during vacuum annealing at temperatures up to 1600 °C was determined. Fairly strong silver diffusion was observed after an initial 10 h annealing period at 1300 °C in both polycrystalline and single crystalline SiC, which is mainly due to implant induced radiation damage. After further annealing at this temperature no additional diffusion took place in the 6H-SiC samples, while it was considerably reduced in the CVD-SiC. The latter was obviously due to grain boundary diffusion and could be described by the Fick diffusion equation. Isochronal annealing of CVD-SiC up to 1400 °C exhibited an Arrhenius type temperature dependence, from which a frequency factor D_o ∼ 4 × 10⁻¹² m² s⁻¹ and an activation energy E_a ∼ 4 × 10⁻¹⁹ J could be extracted. Annealing of 6H-SiC above 1400 °C shifted the silver profile without any broadening towards the surface, where most of the silver was released at 1600 °C. Electron microscopy revealed that this process was accompanied by significant re-structuring of the surface region. An upper limit of D < 10⁻²¹ m² s⁻¹ was estimated for 6H-SiC at 1300 °C.     

Enhanced diffusion under alpha self-irradiation in spent nuclear fuel: Theoretical approaches

Various theoretical approaches have been developed in order to estimate the enhanced diffusion coefficient of fission products under alpha self-irradiation in spent nuclear fuel. These simplified models calculate the effects of alpha particles and recoil atoms on mobility of uranium atoms in UO₂. They lead to a diffusion coefficient which is proportional to the volume alpha activity with a proportionality factor of about 10⁻⁴⁴ (m⁵). However, the same models applied for fission lead to a radiation-enhanced diffusion coefficient which is approximately two orders of magnitude lower than values reported in literature for U and Pu. Other models are based on an extrapolation of radiation-enhanced diffusion measured either in reactors or under heavy ion bombardment. These models lead to a proportionality factor between the alpha self-irradiation enhanced diffusion coefficient and the volume alpha activity of 2 × 10⁻⁴¹ (m⁵).     

Effects of dissolved oxygen on electrochemical and semiconductor properties of 316L stainless steel

The effects of dissolved oxygen on the electrochemical behavior and semiconductor properties of passive film formed on 316L SS in three solutions with different dissolved oxygen were studied by using polarization curve, Mott-Schottky analysis and the point defect model (PDM). The results show that higher dissolved oxygen accelerates both anodic and cathodic process. Based on Mott-Schottky analysis and PDM, the key parameters for passive film, donor density N_d, flat-band potential E_fb and diffusivity of defects D₀ were calculated. The results display that N_d(1−7 × 10²⁷ m⁻³) and D₀(1−18 × 10⁻¹⁶ cm²/s) increase and E_fb value reduces with the dissolved oxygen in solution.     

Frequency dependence studies on the interface trap density and series resistance of HfO2 gate dielectric deposited on Si substrate: Before and after 50 MeV Li ions irradiation

We report the first investigation of the frequency dependent effect of 50 MeV Li³⁺ ion irradiation on the series resistance and interface state density determined from capacitance-voltage (C-V) and conductance-voltage (G-V) characteristics in HfO₂ based MOS capacitors prepared by rf-sputtering. The samples were irradiated by 50 MeV Li³⁺ ions at room temperature. The measured capacitance and conductance were corrected for series resistance. The series resistance was estimated at various frequencies from 1 KHz to 1 MHz before and after irradiation. It was observed that the series resistance decreases from 6344.5 to 322 Ω as a function of frequency before irradiation and 8954-134 Ω after irradiation. The interface state density D_it decreases from 1.12 × 10¹² eV⁻¹ cm⁻² before irradiation to 3.67 × 10¹¹ eV⁻¹ cm⁻² after ion irradiation and further decreases with increasing frequency.     

Deuterium retention in plasma spray tungsten coatings exposed to low-energy, high flux D plasma

Two types of porous plasma spray tungsten coatings deposited onto stainless steel and graphite substrates were exposed to low-energy (76 eV ), high-flux (10²² D/m² s) D plasma to ion fluences of (3-4) × 10²⁶ D/m² at various temperatures. Deuterium retention in the W coatings was examined by thermal desorption spectroscopy and the D(³He,p)⁴He nuclear reaction, allowing determination of the D concentration at depths up to 7 μm. The relatively high D concentration (above 0.1 at.%) at depths of several micrometers observed after D plasma exposure at 340-560 K can be related to accumulation of D₂ molecules in pores, while at temperatures above 600 K deuterium is accumulated mainly in the form of D atoms chemisorbed on the inner pore surfaces. At exposure temperatures above 500 K, the D retention in the plasma spray W coating on graphite substrate increases significantly due to trapping of diffusing D atoms at carbon dangling bonds located at the edge of a graphite crystallite.