We measured neutron-induced reaction cross-sections for Y(n,γ)Y and Y(n,α)Rb reactions with the average neutron energy region from 7.45 to 24.83 MeV by an activation and off-line γ-ray spectrometric technique using the MC-50 Cyclotron at Korea Institute of Radiological and Medical Sciences. The neutron-induced reaction cross-sections of Y as a function of neutron energy were taken from the TEND...

The accuracy of the results obtained using the surface source write/read feature of MCNP for toroidal fusion systems was assessed. Techniques that enhance the effectiveness of the surface source method were identified. Accurate results can be obtained for a specific component in a fusion system by either inserting the detailed component CAD geometry in the full model for the fusion system or ut...

Since 2005 the nodal diffusion based code system, OSCAR-3, was used for reactor support calculations of operational cycles High Flux Reactor in Petten, The Netherlands. OSCAR uses a two-step deterministic calculation, which homogenized cross sections are generated lattice environments using neutron transport simulations, and then passed to core simulator model full reactor. Limitations OSCAR-3 ...

In this study, a series of barium titanate ceramics the chemical composition BaTiO3 was prepared. The solid-state reaction route adopted to synthesize ceramic samples at various sintering temperatures 1100–1300 °C. X-ray diffraction and FTIR spectroscopy were utilized examine structure fabricated ceramics. UV–Vis–reflectance data recorded guess optical bandgap energy synthesized ability attenua...

Background: In CT systems, the machine utilizes a bowtie filter to shape the X-ray beam and remove lower energy photons. The shape of this bowtie filter is complicated and its geometry is often not available in detail. These renders the CT dose index (CTDI) to have different values in measurement versus Monte Carlo simulation studies and other analytical calculations especially in dosimetry of ...

The use of germanium detectors for the identification and quantification of radionuclides in unknown and non-standard counting geometries has increased in the recent past with the need to clean up and verify sites used for radionuclide processing, smuggling detection and other purposes. The calculation of the amount of radionuclide present requires a knowledge of the efficiency of the detector ...