Proton backscattering as a technique for light ion surface interaction studies in CTR materials investigations

Analysis of near-surface helium and hydrogen isotope depth profiles by a specialized proton backscattering technique has been demonstrated as a means to characterize the concentration, initial position, and subsequent migration behavior of all $\text{Z}$ atom species (except ¹H) in metals. The projected range $\text{(R}{}_{\mathrm{<mtext>p</mtext>}}\text{)}$ of helium implanted at 50 keV in Nb, V Ti and Cu has been measured and, for the latter two metals, has been found to agree with theoretical calculations within 100 Å (3%). Deuterium has also been detected and profiled (in Ti) using this technique. Detection sensitivity has been demonstrated at the 7 at % level for D in Ti and at the 0.5 at % level for ⁴He in Cu. In addition, surface and bulk distributions of carbon, oxygen and ³He have been profiled and, in principle, the depth distributions of tritium, ⁶Li, ⁷Li, Be, B, N, and F are also simultaneously resolvable if contained within the foil samples. The technique has been used to investigate the effect of radiation damage and in situ annealing on implanted helium profiles in copper. From these observations the mechanism for helium release from the end-of-range location has been delineated.     

Quasi-thermodynamic prediction of hydrogen reemission behavior from titanium films

A new approach is developed to describe hydrogen reemission behavior from a reactive metal film in which the implanted hydrogen concentration exceeds the dilute solution range. In this approach, conventional thermodynamic information for the metal-hydrogen system, namely, the phase diagram and the $\text{P-C-T}$ (pressure-concentration-temperature) relation is used extensively to determine the reemitted hydrogen flux. As a specific case, this method has been applied to hydrogen reemission from a titanium film. The predicted reemission curves show characteristic discontinuities in time corresponding to phase changes in the solid titanium-hydrogen system as the system proceeds to a final stationary state. Such discontinuities have been found in existing experimental data on deuterium reemission from titanium oxide and carbide coatings but were neglected as being exceptional. The corresponding hydrogen concentration build-up in the titanium film does smoothly approach the stationary state value. The effects of temperature (300–800 K), titanium film thickness (10–3000 Å), and incoming hydrogen flux ($\text{1.0 × 10}{}^{16}$ to $\text{1.0 × 10}{}^{18}$ H atoms/cm².s) on the hydrogen reemission behavior are investigated. The time required to reach a stationary state changes reasonably with incoming hydrogen flux and film thickness. This characteristic transient time minimizes at ~ 600 K and maximizes at ~ 700 K. The stationary state hydrogen concentration is found to be independent of film thickness. An $\text{F-C-T}$ (flux-concentration-temperature) relation is derived and used to estimate the stationary state hydrogen concentration for a given flux and temperature.     

Determination of tritium using a small Van de Graaff accelerator

Determination of $\text{T}\text{D}$ ratios in heavy water samples were carried out using a small Van de Graaff accelerator. The accelerator mass spectrometry system used is described. For comparison, the $\text{T}\text{D}$ ratios were also determined with beta-ray counting methods. Our result demonstrates that tritium at a concentration of 1 part in 10¹⁴ of deuterium was determined with a small Van de Graaff accelerator, and the potential sensitivity would be 2–3 orders of magnitude higher. A simple and cheap 1–2 MeV small accelerator with a suitable ion source and high intensity ion beam may be a powerful tool for determination of tritium.     

First wall surface problems for a D ↽ T tokamak reactor

The effects of deuterium, tritium, helium and neutron bombardment on surface degradation of the first wall of a 5000 MW_th D-T reactor have been analyzed. The effects of both sputtering and blistering have been analyzed and the results applied to 316 stainless steel wall operating at temperatures from 300 to 500°C. It has been calculated that the total wall erosion rate is 0.22 mm/year and that 14 MeV neutron sputtering accounts for two thirds of this number. Sputtering from all neutrons results in $\text{≈0.17}\text{mm/year}$ erosion. The calculated erosion rate is 2–3 times that which would be allowable for a 30 year first wall lifetime.     

Decomposition pressures in the (α + β) fields of the Li-LiH,Li-LiD,and Li-LiT systems

Decomposition (“plateau”) pressures of H₂, D₂, and T₂ over Li-LiH, Li-LiD, and Li-LiT alloys were determined between 600 and 850°C, for pressures ranging from 3 to 460 Torr, and for alloy compositions falling within the (α + β) miscibility gaps. The measurements were carried out separately for each hydrogen isotope using the same lithium sample and experimental procedures. For each system the ln P vs. 1/T data form a pair of linear segments, the intersection of which represents the monotectic temperature (694°C for Li-LiH, 690°C for Li-LiD, and 688°C for Li-LiT). For a given temperature plateau pressures were in the order P_T2 >P_D2 >P_H2. The $\text{D}\text{H}$ and $\text{T}\text{H}$ isotope effects in pressures varied from 1.48 and 1.74 at 600°C to 1.34 and 1.45 at 850°C. The results were used to calculate the standard free energies of formation of solid LiH, LiD, and LiT. The tritium gas used in this study had significant amounts of hydrogen and deuterium. A method for correcting the plateau pressures of this mixture to those of pure tritium is presented.     

The onset of radiation blistering as a function of target temperature for D+ irradiated copper

Radiation blistering in 200 keV D⁺ irradiated copper has been studied for target temperatures in the range 120 to 380 K. Abrupt changes in both the critical dose for blistering and the blister morphology are observed to occur at target temperatures of 195 ± 10 and 300 ± 5 K. For target temperatures in the ranges 120 to 190 K and 300 to 380 K, semi-spherical blisters with a mean diameter of ~5.5 μm are observed to form at fluences of $\text{(2.0 ± 0.5) × 10}{}^{22}$ and $\text{(4 ± 1) × 10}{}^{22}\text{d/m}{}^2$, respectively. In the intermediate temperature region, irregularly shaped blisters with a mean diameter of ~1.3 μm are found to form at a fluence of $\text{(6 ± 2) × 10}{}^{21}\text{d/m}{}^2$. Depth profiles of the implanted deuterium have been obtained. The deuterium concentrations at the onset of blistering are found to be approximately constant over the temperature ranges 120 to 190 K and 190 to 300 K, although a different concentration is observed in each range. For temperatures above 300 K, the deuterium concentration at the onset of blistering is found to decrease with increasing temperature.     

Zum verhalten von tritium in reaktorgraphiten

An experimental investigation of the behaviour of tritium and hydrogen in HTR graphites is described. The diffusion and adsorption kinetic isotherms were measured using a specially developed vacuum apparatus with a variable temperature/time function. Hydrogen was estimated using a mass spectrometer. Tritium was analysed quantitatively after catalytic oxidation using a liquid scintillator. The diffusion coefficients for tritium in reactor graphite were found to be as follows: $\text{D = 0.024 cm}{}^2\text{/sec}\text{exp}\text{(?2.78 eV/kT)}$ and the desorption energies $\text{E}{}^{\mathrm{I}}{}_{\mathrm{des}}\text{= 1.38 ± 0.07 eV}$, $\text{E}{}^{\mathrm{II}}{}_{\mathrm{des}}\text{= 2.07 ± 0.13 eV}$, $\text{E}{}^{\mathrm{III}}{}_{\mathrm{des}}\text{= 3.27 ± 0.07 eV}$. The results indicated that the actual tritium inventory is largely determined by the buffering action of the graphite core. In the start-up phase of an HTR, the tritium concentration in the coolant gas remains relatively low, despite an increased rate of production because the tritium is being adsorbed by the graphite. In later stages of operation, after equilibrium has been reached, the tritium level will be supplemented by material from the hot core regions, so that the coolant activity is not determined by the production rate alone.     

Implantation profiles and sputtering studied by detecting the optical radiation from sputtered particles during ion bombardment

A method of sputter-etching and simultaneous registration of the emitted light is used to obtain depth profiles of implanted Li atoms. Measured mean penetration depths (R_p) and straggling values (ΔR_p) for implanted Li into Ag, V, Si and Al are reported. We find e.g. for 10 keV Li⁺ into $\text{V: R}{}_{\mathrm{n}}\text{= 400 ± 30}\text{A}\text{?}$ and $\text{ΔR}{}_{\mathrm{p}}\text{= 250 ± 30}\text{A}\text{?}$. Known implanted profiles are used as a scale of depth to determine sputtering yields of keV He⁺ ions. With this new method the sputtering yield of 40 keV He⁺ ions bombarding Ag was found to be 0.095 ± 0.020. Sputtering through a thin film is also used to determine sputtering yields. A remarkable increase in the light intensity from excited Ag I atoms is observed during sputtering through the interface between an Ag film and the underlying Al substrate. This is found to be due to a change in the excitation mechanism. Continuous features in the observed emission spectrum have in a few cases been identified as originating from deexcitation radiation of molecules formed in the sputtering process.     

Void formation and solute segragation in Ti-14.4 at% Al

An alloy of Ti-14.4 at% Al was irradiated with 3.0 MeV ⁵⁸Ni ions at temperatures near 600°C and to damage levels from 1.3 to 14.7 dpa. TEM examination of these specimens showed the damage produced voids, dislocation loops and network dislocations. The undersize aluminum solute atom segregated to the voids, grain boundaries and the irradiated surface, causing enhanced precipitation of γ'-Ti₃Al at these sinks for point defects. Shells of $\text{α}{}_2\text{~ 75}\text{A}\text{?}$ thick coated the voids and a continuous film of γ' was observed on the irradiated surface and on some grain boundaries. The surface film of α₂ contained $\text{1}\text{3}\text{a 〈11}\text{2}\text{?}\text{0〉}$ antiphase boundaries. Voids have not been previously observed in irradiated hcp Ti and this is the first observation of substantial irradiation-induced segregation in an hcp alloy.     

Shear-like transformation in β-stabilized U-1.5 at% Ga alloy; structure of the intermetallic compound U3Ga5

By rapid cooling, the β-U structure can be retained at room temperature in an U-1.5 at% Ga alloy. The metastable structure subsequently undergoes an isothermal shear-like transformation into a supersaturated solid solution of gallium in α-uranium. The intermetallic compound with the highest uranium content in the binary U-Ga system is U₃Ga₅, space group Cmcm. Pu₃Pd₅-type, with lattice parameters: $\text{a = 9.396}$, $\text{b = 7.575}$ and $\text{c = 9.387}$ Å.     

Depth distribution profiling of deuterium and 3He

A technique using nuclear microanalysis has been developed to determine ²D an ³He concentrations versus depth profiles in the near surface regions (～2 μm) of solids. Ultimate near-surface depth resolutions of (≈100 Å) should be attainable by these techniques. A probing beam of ³He ions is used to analyze for the deuterium. By energy analyzing the emitted ⁴He ions from the reaction d(³He,p)⁴He and by simple computer analysis, deuterium concentration versus depth profiles are obtained. The same methods are applicable for ³He profiling except in that case a ²D probing beam is used. The techniques is currently being developed using Er hydride and Sc hydride films. It is particularly suited for the study of the transition metals with high hydrogen solubility such as Nb and V which are of interest for CTR reactor applications.     

Irradiation effects of boron carbide used as control rod elements in fast breeder reactors

The paper describes radiation effects on 8₄C pellets used as control rod elements in the Enrico Fermi Fast Breeder Reactor. Pellet swelling (Δ$\text{V/V}$) caused by irradiation was less than 1% in which crystal lattice swelling was less than 20%. Many microcracks, a main cause of pellet swelling, appeared in the irradiated pellets. The production of microcracks was related to graphite precipitation in the pellets before irradiation. Open pores which did not exist in the unirradiated pellets were formed in the irradiated ones. In a unit cell of B₄C, the α-axis elongated by 0.025 Å and the $\text{c-}\text{axis}$ shrank by 0.07 Å by irradiation. Moreover, we found three recovery stages which were from room temperature to 400°C, from 400 to 750°C and from 850 to 1100°C. The recovery mechanisms in the irradiated pellets are discussed in terms of the helium behavior.     

Comparison of 16 MeV proton, 14 MeV neutron and fission neutron damage in copper

The irradiation damage structures produced in high-purity copper by a fluence of $\text{3 × 10}{}^{16}$ particles/cm² of 16 MeV protons, 14 MeV neutrons, and fission neutrons ($\text{E > 1 MeV}$) were studied by transmission electron microscopy. The damage consists of vacancy-and-interstitial clusters or sessile Frank dislocation loops oriented on {111} planes of the copper matrix, and ranges in size from 25 Å (lower limit of resolution) to 200Å in diameter. p]The size-density distributions of the clusters in the 14 MeV neutron and 16 MeV proton irradiated samples were virtually identical, and the average size of the clusters in these two groups of specimens was substantially larger than was the case for those in the fission-neutron-irradiated copper.     

Chemical states and thermal stability of hydrogen-implanted Ti and V studied by X-ray photoelectron spectroscopy

X-ray photoelectron spectroscopy (XPS) was applied to the chemical state studies of Ti and V metals bombarded with 8 keV hydrogen ions. The binding energies of the lines for the ion-implanted Ti and V shift from those for the metallic states by 0.3 eV, are consistent with the core-line shifts for the thermally synthesized hydrides such as TiH_1.97 and VH_0.55 The metal 3d-H1s bonding level for the hydrogen implanted Ti appears at ~ 3.5 eV below the Fermi level, which is lower by ~ 2.5 eV than the molecular-orbital energy previously calculated for TiH₂. In the case of the ion-implanted V, however, it appears at ~ 5.0 eV, which is almost equal to the molecular-orbital energy for VH₂. The photopeaks corresponding to the Ti-H and V-H bonds for the ion-implanted samples grew up on raising the annealing temperature up to 550°C and 150°C, respectively. The phenomena are interpreted by means of the thermal diffusion of the implanted hydrogen from bulk to surface.     

Tritium solubility in SUS-316 stainless steel

Tritium solubility in SUS-316 stainless steel was determined with a gas absorption method, in which tritium gas diluted by protium was used. The tritium absorption experiments were carried out at temperatures of 703, 804 and 903 K under pressures of 10, 30, 50 and 100 torr of tritiated hydrogen gas. The radioactivity of tritium dissolved in the specimen was measured by the method of liquid scintillation counting.The tritium solubility was derived from the experimental data by taking into consideration of isotopic equilibrium among H₂, T₂ and HT molecules. The determined tritium solubility can be expressed by the equation:

     

Lattice diffusion coefficient of oxygen in lithium oxide

Lattice diffusion coefficients of the oxygen ion in antifluorite-cubic Li₂O were determined employing polycrystalline samples, in the temperature range of 920 ~ 1130°C, by means of the gas-solid isotope exchange and solid-phase analysis technique; use was made of the relationship between the particle-size and grain-size dependences of the grain-boundary enhanced diffusion. The results were described by D = 1.52 × 10³ exp$\text{(?83.3 × 10}{}^3\text{/RT)}\text{cm}{}^2\text{s}$ and showed a good agreement with the previously reported results determined by the gas-phase analysis technique. Diffusion characteristics of the constituent ions in the antifluorite-cubic structure were discussed in comparison with those in fluorite-cubic crystals.     

Study on the phase behavior of CayU1−yO2+x solid solution

The lattice parameter and O/M ratio of the solid solution, Ca_yU_1?yO_2+x, prepared in a stream of helium were investigated. The single phase region of the cubic fluorite type solid solution extended from $\text{0.03 ? y ? 0.33}$. The lattice parameter was expressed as a linear equation of $\text{x}$ and $\text{y}$ in the range $\text{x ? 0}$ to be $\text{a = 5.4704 ? 0.102x ? 0.310y}$ (Å).The oxidation states of uranium in the solid solution were found to be U⁴⁺ and U⁵⁺ from a simple ionic model assuming rigid sphere ions. The partial molar enthalpies were estimated using calculated partial molar entropy. The partial molar enthalpy showed a maximum near $\text{x = 0.01}$.     

Infrared spectroscopic analysis of OH− and OD− in crystalline Li2O as a function of chemical treatment

Infrared spectroscopy has been used to study the chemical form and approximate concentration of OH^? and OD^? in Li₂O single crystals as a function of chemical treatment. Infrared absorption maxima at $\text{(3671±0.5)}\text{cm}{}^{\mathrm{?1}}\text{and}\text{(2711±3.3)}\text{cm}{}^{\mathrm{?1}}$ were observed for OH^? and OD^?, respectively. The absorption coefficient for OD^? was determined to be $\text{4.0±0.4}$ absorbance units per mol part per million OD^? per mm of sample thickness. Vacuum baking of Li₂O crystals reduced the OH^? and OD^? concentrations to <50 mppm; baking in a low moisture-level D₂ environment at 600 to 800°C appeared to lead to volatilization of LiD from the Li₂O crystals; and baking in D₂ containing ($\text{350±50}$) mppm D₂O at 600 to 800°C produced a measurable quantity of LiOD. In all cases, the observed spectra indicated the presence of only one distinguishable form of OH^? or OD^? in the Li₂O lattice. Because of the close correspondence of the observed absorption maxima to reported values for pure LiOH and LiOD, the most consistent (although not conclusive) interpretation is that the OH^? and OD^? are present as a separate LiOH or LiOD phase at room temperature. Only limited conclusions can be drawn regarding the chemical state of OH^? and OD^? during the elevated temperatures exposures. An estimate of the approximate value for the solubility of tritium in Li₂O at 800°C was made using data from D₂/Li₂O isothermal exposure experiments — this value was ? 25 wppm.     

Untersuchungen zur rückhaltung von tritium in HTR-brennelementen

In a nuclear reactor tritium is produced by ternary fission and a number of further neutron-induced nuclear reactions. The liberation of the tritium so produced from the ceramic fuel of a high temperature reactor was examined. Fuel particles with pyrocarbon coatings are able to retain tritium to a large extent under reactor operating conditions. The amounts of tritium found in reactorirradiated fuel particles correspond within the limits of error with the calculated concentrations. The activation energy of the liberation factor for tritium from fuel particles is 106 kcal/mol in the temperature range 1400–1900°C. The release of the tritium found according to ${}^6\text{Li (n, α)}{}^3\text{H}$ from a reactor graphite was measured and the concentration of tritium in graphite samples from the AVR nuclear power station was determined.