TiN-based coatings on fuel cladding tubes for advanced nuclear reactors

Titanium nitride (TiN) thin films are coated on HT-9 and MA957 fuel cladding tubes and bars to explore their mechanical strength, thermal stability, diffusion barrier properties, and thermal conductivity properties. The ultimate goal is to implement TiN as an effective diffusion barrier to prevent the inter-diffusion between the nuclear fuel and the cladding material, and thus lead to a longer lifetime of the cladding tubes. Mechanical tests including hardness and scratch tests for the samples before and after thermal cycle tests show that the films have a high hardness of 28GPa and excellent adhesion properties despite the thermal treatment. Thermal conductivity measurements demonstrate that the thin TiN films have very minimal impact on the overall thermal conductivity of the MA957 and HT-9 substrates, i.e., the thermal conductivity of the uncoated HT-9 and MA957 substrates was 26.25 and 28.44 W m-1 K-1, and that of the coated ones was 26.21 and 28.38W m-1 K-1, respectively. A preliminary Ce diffusion test on the couple of Ce/TiN/HT-9 suggests that TiN has excellent material compatibility and good diffusion barrier properties.

An exploratory study to determine applicability of nano-hardness and micro-compression measurements for yield stress estimation

The superior properties of ferritic/martensitic steels in a radiation environment (low swelling, low activation under irradiation and good corrosion resistance) make them good candidates for structural parts in future reactors and spallation sources. While it cannot substitute for true reactor experiments, irradiation by charged particles from accelerators can reduce the number of reactor experiments and support fundamental research for a better understanding of radiation effects in materials. Based on the nature of low energy accelerator experiments, only a small volume of material can be uniformly irradiated. Micro and nanoscale post irradiation tests thus have to be performed. We show here that nanoindentation and micro-compression testing on T91 and HT-9 stainless steel before and after ion irradiation are useful methods to evaluate the radiation induced hardening.

Oxygen effects on irradiated tantalum alloys

Ta and Ta-1% W are being considered to be used as target clad materials in the LANSCE proton beam line for the material test station (MTS). To investigate the embrittlement of these materials due to oxygen contamination and proton irradiation, Ta and Ta-1 wt% W (as received and with ~400 ppm O) were exposed to a 3.5 MeV proton beam at the ion beam materials laboratory at LANL. After irradiating the samples in the proton beam, nanoindentation was performed in cross-section to investigate the hardness increase of the materials due to irradiation. The nanoindentation showed that the hardness increase due to irradiation is between 9% and 20% depending on the material. The results show good agreement with mechanical testing results on tantalum and Ta-1 wt% W after high energy proton irradiation to doses up to 23 dpa.

Nano-indentation measurement of oxide layers formed in LBE on F/M steels

Ferritic/martensitic (F/M) steels (T91, HT-9, EP 823) are candidate materials for future liquid lead or lead bismuth eutectic (LBE) cooled nuclear reactors. To understand the corrosion of these materials in LBE, samples of each material were exposed at 535 °C for 600 h and 200 h at an oxygen content of 10 wt%. After the corrosion tests, the samples were analyzed using SEM, WDX and nano-indentation in cross section. Multi-layered oxide scales were found on the sample surfaces. The compositions of these oxide layers are not entirely in agreement with the literature. The nano-indentation results showed that the E-modulus and hardness of the oxide layers are significantly lower than the values for dense bulk oxide materials. It is assumed that the low values stem from high porosity in the oxide layers. Comparison with in-air oxidized steels show that the E-modulus decreases with increasing oxide layer thickness. © 2008 Elsevier B.V. All rights reserved.

Size dependent enhancement of helium ion irradiation tolerance in sputtered Cu/V nanolaminates

We have investigated the evolution of radiation damage and changes in hardness of sputter-deposited Cu/V nanolaminates upon room temperature helium ion irradiation. As the individual layer thickness decreases from 200 to 5 nm, helium bubble density and radiation hardening both decrease. The magnitude of radiation hardening becomes negligible for individual layer thickness of 2.5 nm or less. These observations indicate that nearly immiscible Cu/V interface can effectively absorb radiation-induced point defects and reduce their concentrations.

Characterization of oxide layers grown on D9 austenitic stainless steel in lead bismuth eutectic

Lead bismuth eutectic (LBE) is a possible coolant for fast reactors and targets in spallation neutron sources. Its low melting point, high evaporation point, good thermal conductivity, low reactivity, and good neutron yield make it a safe and high performance coolant in radiation environments. The disadvantage is that it is a corrosive medium for most steels and container materials. This study was performed to evaluate the corrosion behavior of the austenitic stainless steel D9 in oxygen controlled LBE. In order to predict the corrosion behavior of steel in this environment detailed analyses have to be performed on the oxide layers formed on these materials and various other relevant materials upon exposure to LBE. In this study the corrosion/oxidation of D9 stainless steel in LBE was investigated in great detail. The oxide layers formed were characterized using atomic force microscopy, magnetic force microscopy, nanoindentation, and scanning electron microscopy with wavelength-dispersive spectroscopy (WDS) to understand the corrosion and oxidation mechanisms of D9 stainless steel in contact with the LBE. What was previously believed to be a simple double oxide layer was identified here to consist of at least 4 different oxide layers. It was found that the inner most oxide layer takes over the grain structure of what used to be the bulk steel material while the outer oxide layer consists of freshly grown oxides with a columnar structure. These results lead to a descriptive model of how these oxide layers grow on this steel under the harsh environments encountered in these applications.

Size-dependent radiation tolerance in ion irradiated TiN/AlN nanolayer films

Interface effects on ion-irradiation tolerance properties are investigated in nanolayered TiN/AlN films with individual layer thickness varied from 5 nm to 50 nm, prepared by pulsed laser deposition. Evolution of the microstructure and hardness of the multilayer films are examined on the specimens before and after He ion-implantation to a fluence of 4 × 10 m at 50 keV. The suppression of amorphization in AlN layers and the reduction of radiation-induced softening are observed in all nanolayer films. A clear size-dependent radiation tolerance characteristic is observed in the nanolayer films, i.e., the samples with the optimum layer thickness from 10 nm to 20 nm show the best ion irradiation tolerance properties, and a critical layer thickness of more than 5 nm is necessary to prevent severe intermixing. This study suggests that both the interface characteristics and the critical length scale (layer thickness) contribute to the reduction of the radiation-induced damages in nitride-based ceramic materials. © 2013 Elsevier B.V. All rights reserved.

The interaction of fast neutrons with shielding and fusion blanket materials

In the present work the neutron emission spectra from a graphite cube, and from natural uranium, lithium fluoride, graphite, lead and steel slabs bombarded with 14.1 MeV neutrons were measured to test nuclear data and calculational methods for D - T fusion reactor neutronics. The neutron spectra measured were performed by an organic scintillator using a pulse shape discrimination technique based on a charge comparison method to reject the gamma rays counts. A computer programme was used to analyse the experimental data by the differentiation unfolding method. The 14.1 MeV neutron source was obtained from T(d,n)4He reaction by the bombardment of T - Ti target with a deuteron beam of energy 130 KeV. The total neutron yield was monitored by the associated particle method using a silicon surface barrier detector. The numerical calculations were performed using the one-dimensional discrete-ordinate neutron transport code ANISN with the ZZ-FEWG 1/ 31-1F cross section library. A computer programme based on Gaussian smoothing function was used to smooth the calculated data and to match the experimental data. There was general agreement between measured and calculated spectra for the range of materials studied. The ANISN calculations carried out with P3 - S8 calculations together with representation of the slab assemblies by a hollow sphere with no reflection at the internal boundary were adequate to model the experimental data and hence it appears that the cross section set is satisfactory and for the materials tested needs no modification in the range 14.1 MeV to 2 MeV. Also it would be possible to carry out a study on fusion reactor blankets, using cylindrical geometry and including a series of concentric cylindrical shells to represent the torus wall, possible neutron converter and breeder regions, and reflector and shielding regions.