Irradiation growth during fission fragment and proton bombardment of uranium

Irradiation growth results are reported for annealed -uranium at 373 K under 3.5 MeV proton bombardment. Two such experiments were performed at damage rates of 6.9 × 10⁻⁸ and 9.3 × 10⁻⁸ dpa/s to doses of 0.0072 and 0.0077 dpa, respectively. In each case the growth rate remained constant throughout the experiment. The respective damage normalised growth rates were 5.6 × 10⁻³ and 7.1 × 10⁻³ dpa⁻¹. Comparison between proton growth rates and published in-reactor growth rates is made by converting the more usual fuel damage parameters, such as burn-up, to dpa. Damage calculations, using the NRT damage model, are presented which indicate that, in uranium, each fission event produces 100 000 displacements. The reported growth rate of annealed, polycrystalline -uranium at 353 K, during thermal neutron irradiation, represents a damage normalised growth rate of 9.6 × 10⁻³ dpa⁻¹, which is not substantially different from the present proton results. This similarity of proton and fission growth rates appears to be contrary to the earlier finding of Thompson (1960), who deduced that proton bombardment produced two orders of magnitude less growth than fission fragments. Thompson concluded that thermal spikes played a dominant role in irradiation growth. Thompson's results and analysis are reassessed in the light of recent range data and damage models and found to be consistent with the present results in both magnitude and direction. The results are also inconsistent with Buckley's original model to the extent that thermal spikes were thought to play an important role. From a consideration of primary recoil spectra it is shown that the concept of the anisotropic aggregation of point defects to form vacancy and interstitial clusters, which is at the centre of that model, remains viable. Furthermore, similar though slightly less growth would be expected during proton bombardment. This was indeed found to be the case, the growth rate with protons being about half that with fission fragments.     

Low temperature neutron irradiation effects on microstructure and tensile properties of molybdenum

Polycrystalline molybdenum was irradiated in the hydraulic tube facility at the High Flux Isotope Reactor to doses ranging from 7.2 × 10⁻⁵ to 0.28 dpa at 80 °C. As-irradiated microstructure was characterized by room-temperature electrical resistivity measurements, transmission electron microscopy (TEM) and positron annihilation spectroscopy (PAS). Tensile tests were carried out between −50 and 100 °C over the strain rate range 1 × 10⁻⁵ to 1 × 10⁻² s⁻¹. Fractography was performed by scanning electron microscopy (SEM), and the deformation microstructure was examined by TEM after tensile testing. Irradiation-induced defects became visible by TEM at 0.001 dpa. Both their density and mean size increased with increasing dose. Submicroscopic three-dimensional cavities were detected by PAS even at 0.0001 dpa. The cavity density increased with increasing dose, while their mean size and size distribution was relatively insensitive to neutron dose. It is suggested that the formation of visible dislocation loops was predominantly a nucleation and growth process, while in-cascade vacancy clustering may be significant in Mo. Neutron irradiation reduced the temperature and strain rate dependence of the yield stress, leading to radiation softening in Mo at lower doses. Irradiation had practically no influence on the magnitude and the temperature and strain rate dependence of the plastic instability stress.     

Angular effects in the sputtering of Cr from stainless steel by 10 keV H3

Energy and angular distributions of Cr⁺ sputtered from stainless steel by 1.6 × 10⁻¹⁵ J (10 keV) H⁺₃ are reported as a function of angle of incidence. For more normal incidence, the peak in the energy distribution occurs in the vicinity of 3.2 × 10⁻¹⁹ J (2 eV), the average energy is approximately 1.12 × 10⁻¹⁸ J (7 eV), and the angular distribution is close to cosine. Toward glancing incidence, the peak energy increases to ˜6.4 × 10⁻¹⁹ J (4 eV), the average energy increases to ˜1.28 × 10⁻¹⁸ J (8.0 eV), and the angular distribution shows a distinct maximum in the forward direction. These results are discussed in terms of the increasing role of surface recoils in the sputtering mechanism at glancing incidence.     

In-reactor uniaxial tensile testing of pure copper at a constant strain rate at 90 °C

Recently, annealed specimens of pure copper have been tensile tested in a fission reactor at a damage rate of 6 × 10⁻⁸ dpa/s with a constant strain rate of 1.3 × 10⁻⁷ s⁻¹. The specimen temperature during the test was about 90 °C. The stress response was continuously recorded as a function of irradiation time (i.e. displacement dose and strain). The experiment lasted for 308 h. During the dynamic in-reactor test, the specimen deformed and hardened homogeneously without showing any sign of yield drop and plastic instability. However, the specimen yielded a uniform elongation of only about 12%. The preliminary results are briefly described and discussed in the present note.     

Mechanism of dynamic strain aging and characterization of its effect on the low-cycle fatigue behavior in type 316L stainless steel

Low-cycle fatigue tests were carried out in air in a wide temperature range from 20 to 650 °C with strain rates of 3.2 × 10⁻⁵–1 × 10⁻² s⁻¹ for type 316L stainless steel to investigate dynamic strain aging (DSA) effect on the fatigue resistance. The regime of DSA was evaluated using the anomalies associated with DSA and was in the temperature range of 250–550 °C at a strain rate of 1 × 10⁻⁴ s⁻¹, in 250–600 °C at 1 × 10⁻³ s⁻¹, and in 250–650 °C at 1 × 10⁻² s⁻¹. The activation energies for each type of serration were about 0.57–0.74 times those for lattice diffusion indicating that a mechanism other than lattice diffusion is involved. It seems to be reasonable to infer that DSA is caused by the pipe diffusion of solute atoms through the dislocation core. Dynamic strain aging reduced the crack initiation and propagation life by way of multiple crack initiation, which comes from the DSA-induced inhomogeneity of deformation, and rapid crack propagation due to the DSA-induced hardening, respectively.     

Measurements of the erosion of stainless steel, carbon, and SiC by hydrogen bombardment in the energy range of 0.5 – 7.5 keV

Total erosion yields by sputtering and blistering for 1 to 15 keV H₂⁺ bombardment at normal incidence have been measured by weight loss of 304 stainless steel, pyrolytic graphite, carbon fibres, glassy carbon and SiC. The erosion yields are in the range of 3 × 10⁻³ to 2.6 × 10⁻² atoms per incident hydrogen atom. Observation in the scanning electron microscope shows that blisters occur in stainless steel and SiC at doses of 5 × 10¹⁸ particles/cm², but disappear at doses of 5 × 10 particles/cm² . The surface roughening observed depends largely on grain orientation. On carbon no blistering could be found. After bombardment the carbon surfaces are generally more smooth than before.     

Development of the Climb Induced Glide concept to describe in-reactor creep of fcc materials

The Climb Induced Glide model (CIG) for irradiation creep is developed using a plastic flow law which has been successfully applied in the correlation of Type 316 stainless steel rupture data. This model is used to predict the stress and temperature dependence of irradiation creep and the transition from irradiation to thermal creep. The predictions of this model are compared and found to be qualitatively consistent with experimental data and microstructural information. This model allows prediction of deformation behavior covering strain rates from 1 × 10⁻¹³ s⁻¹ to 1 s⁻¹.     

Lattice location and electrical conductivity in ion implanted TiO2 single crystals

Single crystals of TiO₂ (rutile) were implanted at room temperature with Ar, Sn and W ions applying fluences of 10¹⁵/cm² to 10¹⁶/cm² at 300 keV. The lattice location, together with ion range and damage distribution was measured with Rutherford Backscattering and Channeling (RBS-C). The conductivity, σ, was measured as a function of temperature. The implanted Sn and W atoms were entirely substitutional on Ti sites in the applied fluence region, where the radiation damage did not yet reach the random level. A large σ increase was observed for all implants at displacement per atom values (dpa) below 1. Above dpa = 1, σ reveals a saturation value of 0.3 Ω⁻¹ cm⁻¹ for Ar implants, while for W and Sn implants a further increase of σ up to 30 Ω⁻¹ cm⁻¹ was measured. Between 70 K and 293 K ln σ was proportional to T^−1/2, (Ar,W) and T^−1/4 (Sn), indicating that the transport mechanism is due to variable range hopping.     

Measurements of natural concentrations of I in uranium ores by accelerator mass spectrometry

An accelerator mass spectrometry system is described and utilized for measurements of ¹²⁹I concentrations in natural and environmental samples. We report here on measurements of ¹²⁹I isotopic abundances in iodine reagents and in iodine of mineral origin and of ¹²⁹I concentrations in uranium ores of different origins. The ¹²⁹I isotopic abundances for two measured contemporary iodine reagents and for iodine from a deep underground brine are 1.3 × 10⁻¹³ and about 4 × 10⁻¹⁴, respectively. ¹²⁹I/U ratios in the range 10⁻¹³–10⁻¹² are measured and compared to a simple model of ¹²⁹I production by spontaneous and induced fission of uranium. No clear correlation with the uranium concentrations or residence times is observed.     

Isothermal release of tritium from neutron-irradiated Li17Pb83

Isothermal release experiments were carried out to study the tritium recovery from lithium-lead alloy Li₁₇Pb₈₃ in which tritium was produced by irradiation with thermal neutrons. The experimental results indicate that the tritium recovery was incomplete within two hours at 200 °C. At temperatures above the melting point, the tritium release rates have been significantly increased and found to be controlled by the diffusion in the alloy. The determined diffusion coefficients of tritium in the alloy are 6.6 × 10⁻⁶, 7.8 × 10⁻⁶ and 9.5 × 10⁻⁶ cm²/s at 300, 400 and 500°C, respectively.     

The thermal conductivity of UO2 vapor

The thermal conductivity, λ of a saturated vapor over UO_1.96 is calculated in the temperature range 3000–6000 K. The calculation shows that the contribution to λ from the transport of reaction enthalpy dominates all other contributions. All possible reactions of the gaseous species UO₃, UO₂, UO, U, O, and O₂ are included in the calculation. We fit the total thermal conductivity to the empirical equation λ = exp(a+ b/T+cT+dT² + eT³), with λ in cal/(cm s K), T in kelvins, a = 268.90, B = − 3.1919 × 10⁵, C = −8.9673 × 10⁻², d = 1.2861 × 10⁻⁵, and E = −6.7917 × 10⁻¹⁰.     

Time scales of transient enhanced diffusion: free and clustered interstitials

Transient enhanced diffusion (TED) and electrical activation after nonamorphizing Si implantations into lightly B-doped Si multilayers shows two distinct timescales, each related to a different class of interstitial defect. At 700°C, ultrafast TED occurs within the first 15 s with a B diffusivity enhancement of > 2 × 10⁵. Immobile clustered B is present at low concentration levels after the ultrafast transient and persists for an extended period ( 10²–10³ s). The later phase of TED exhibits a near-constant diffusivity enhancement of ≈ 1 × 10⁴, consistent with interstitial injection controlled by dissolving {113} interstitial clusters. The relative contributions of the ultrafast and regular TED regimes to the final diffusive broadening of the B profile depends on the proportion of interstitials that escape capture by {113} clusters growing within the implant damage region upon annealing. Our results explain the ultrafast TED recently observed after medium-dose B implantation. In that case there are enough B atoms to trap a large proportion of interstitials in Si---B clusters, and the remaining interstitials contribute to TED without passing through an intermediate {113} defect stage. The data on the ultrafast TED pulse allows us to extract lower limits for the diffusivities of the Si interstitial (D_I > 2 × 10⁻¹⁰ cm²s⁻¹) and the B interstitial(cy) defect (D_Bi > 2 × 10⁻¹³ cm²s⁻¹) at 700°C.     

Pumping experiment of water on B and LaB6 films with an electron beam evaporator

The pumping characteristic of water vapor on boron and lanthanum hexaboride films formed with an electron beam evaporator have been investigated in high vacuum between 10⁻⁴ and 10⁻³ Pa. The measured initial maximum pumping speeds of water for the fresh B or LaB₆ films with a deposition amount from 2.3 × 10²¹ to 6.7× 10²¹ molecules/m² separately formed on a substrate are 3.2–4.9 m³/sm², and the saturation values of adsorbed water on these films are 2.1 ×10²⁰−1.3 × 10²¹ H₂O molecules/m².     

Modelling the high temperature mechanical behaviour of helium embrittled austenitic stainless steel

Current theories for the growth of grain-boundary defects during plastic deformation have been used to model the ductility of an irradiated austenitic steel as a function of strain rate and thermal-neutron fluence. Failure of irradiated steels is via the growth and linkage of helium bubbles on the grain boundaries produced by an (n, ) reaction between the thermal neutrons and boron present in the steel as a controlled trace element.

Reasonable agreement is obtained between theoretically predicted and experimentally measured ductilities over a range of strain rates between 10⁻¹⁰ and 10⁻²/s. However, in order to obtain such agreement over the range of strain rates examined, variables such as defect size and spacing had to be carefully selected. The values used were internally self-consistent with the amount of helium known to be availablc to nucleate grain-boundary bubbles.

The models of defect growth have also been used to make some predictions on the creep-fatigue performance of such materials.     

Helium retention of vanadium alloy after energetic helium ion irradiation

Helium irradiation experiments of V–4Ti alloy were conducted in an ECR ion irradiation apparatus by using helium ions with energy of 5 keV. The ion fluence was in the range from 1 × 10¹⁷ He/cm² to 8 × 10¹⁷ He/cm². After the helium ion irradiation, the helium retention was examined by using a technique of thermal desorption spectroscopy (TDS). After the irradiation, the blisters with a size of about 0.1 μm were observed at the surface, and the blister density increased with the ion fluence. Two desorption peaks were observed at approximately 500 and 1200 K in the thermal desorption spectrum. When the ion fluence was low, the retained helium desorbed mainly at the higher temperature regime. As increase of the ion fluence, the desorption at the lower temperature peak increased and the retained amount of helium saturated. The saturated amount was approximately 2.5 × 10¹⁷ He/cm². This value was comparable with those of the other plasma facing materials such as graphite.     

Room temperature migration of ion beam injected point defects in crystalline silicon

We have investigated the room temperature diffusion and trapping phenomena of ion beam generated point defects in crystalline Si by monitoring their interaction with dopants, native contaminants such as C and O, and other defects. Spreading resistance measurements show that a small fraction ( 10⁻⁷–10⁻⁶) of the defects generated at the surface by a 40 keV Si implant is injected into the bulk. These defects undergo trap-limited diffusion and produce dopant deactivation and/or partial annihilation of preexisting deep (several micron) defect markers, produced by MeV He implants. It is found that in highly pure, epitaxial Si layers, these effects extend to several microns from the surface, demonstrating a long range migration of point defects at room temperature. A detailed analysis of the experimental evidences allows us to identify the Si self-interstitials injected into the bulk as the major responsible of both dopant deactivation and partial annealing of vacancy-type defects (divacancies, phosphorus-vacancy and oxygen-vacancy) generated by the implants. Finally, a lower limit of 6 × 10⁻¹¹ cm²/s is obtained for the room temperature diffusivity of Si self-interstitials.     

Study of radiation-induced degradation of RPV steels and model alloys by positron annihilation and Mössbauer spectroscopy

The influence of different microstructural processes on the degradation due to radiation embrittlement has studied by positron annihilation and Mössbauer spectroscopy. The materials studied consisted of WWER-440 base (15Kh2MFA) and weld (10KhMFT) RPV steels which were neutron-irradiated at fluence levels of 0.78 × 10²⁴ m⁻², 1.47 × 10²⁴ m⁻² and 2.54 × 10²⁴ m⁻²; WWER-1000 base (15Kh2NMFAA) and weld (12Kh2N2MAA) irradiated at a fluence level 1.12 × 10²⁴ m⁻²; three different model alloys implanted with protons at two dose levels (up to 0.026 dpa), finally the base metal of WWER-1000 (15Kh2NMFAA) was thermally treated with the intention to simulate the P-segregation process. It has been shown possible to correlate the values of parameters obtained by such techniques and data of mechanical testing (ductile-to-brittle transition temperature and upper shelf energy).     

Effect of tin on the irradiation growth of polycrystalline zirconium

Measurements of irradiation growth of polycrystalline Zr-1.5% Sn and Zr-0.1% Sn alloys at 353 K and 553 K have been made following fast neutron irradiation with fluences up to 3.1 × 10²⁵ n/m². At 353 K, growth of Zr-1.5% Sn virtually saturated at a strain of 4.5 × 10⁻⁴ after a fluence of ˜10²⁴ n/m². At this temperature, Zr-0.1% Sn continued to grów until ˜ 2 × 10²⁵ n/m², when the strain levelled off at ˜ 1.2×10⁻³. At 553 K, Zr-1.5% Sn initially grew about twice as fast as the 0.1% Sn alloy, but both eventually reached the same steady state rate of ˜ 2.4 × 10⁻²⁹ m²/n. Comparison of the data for the 1.5% Sn material with those for Zircaloy-2 from earlier work reveals that at 353 K, growth is suppressed by the presence of Sn atoms, which may serve as vacancy traps. However, at 553 K, minor additions and impurities in Zircaloy-2 (such as Fe, Ni, Cr and O) play an important role and cannot be neglected. The growth behaviour of Zr-0.1% Sn is similar to that of pure polycrystalline zirconium, especially at 353 K, indicating that the addition of Sn at this concentration does not strongly influence the growth of zirconium.     

The effects of low dose rate irradiation and thermal aging on reactor structural alloys

As part of the EBR-II reactor materials surveillance program, test samples of fifteen different alloys were placed into EBR-II in 1965. The surveillance (SURV) program was intended to determine property changes in reactor structural materials caused by irradiation and thermal aging. In this work, the effect of low dose rate (approximately 2 × 10⁻⁸ dpa/s) irradiation at 380–410°C and long term thermal aging at 371°C on the properties of 20% cold worked 304 stainless steel, 420 stainless steel, Inconel X750, 304/308 stainless weld material, and 17-4 PH steel are evaluated. Doses of up to 6.8 dpa and thermal aging to 2994 days did not significantly affect the density of these alloys. The strength of 304 SS, X750, 17-4 PH, and 304/308 weld material increased with irradiation. In contrast, the strength of 420 stainless steel decreased with irradiation. Irradiation decreased the impact energy in both Inconel X750 and 17-4 PH steel. Thermal aging decreased the impact energy in 17-4 PH steel and increased the impact energy in Inconel X750. Tensile property comparisons of 304 SURV samples with 304 samples irradiated in EBR-II at a higher dose rate show that the higher dose rate samples had greater increases in strength and greater losses in ductility.     

Effects of ion implantation on poly(dimethylsilylene-co-methylphenylsilylene)

The effects of ion implantation on the electrical and structural properties of poly(dimethylsilylene-co-methylphenylsilylene), (DMMPS) thin films have been investigated. Ionic species of krypton, arsenic, fluorine, chlorine, and sulfur were implanted at energies ranging from 35 to 200 keV and with doses of up to 1 × 10¹⁶ ion cm². The conductivity of the polymer increased upon implantation reaching a maximum value of 9.6 × 10⁻⁶ (Ω cm)⁻¹ for the case of arsenic ion at a dose of 1 × 10¹⁶ ion cm² and energy of 100 keV. The results showed that ion implantation induced conduction in DMMPS was primarily due to structural modifications of the material brought about by the, energetic ions. Infrared analysis and Auger electron spectroscopy showed evidence for the formation of a silicon carbide-like structure upon implantation.