Results 1 - 10 of 75717
Results 1 - 10 of 75717. Search took: 0.05 seconds
|Sort by: date | relevance|
[en] This article presents nuclide-specific organ dose rate coefficients for environmental external exposures due to soil contamination assumed as a planar source at a depth of 0.5 g cm in the soil and submersion to contaminated air, for a pregnant female and its fetus at the 24th week of gestation. Furthermore, air kerma free-in-air coefficient rates are listed. The coefficients relate the organ equivalent dose rates (Sv s) to the activity concentration of environmental sources, in Bq m or Bq m, allowing to time-integrate over a particular exposure period. The environmental radiation fields were simulated with the Monte Carlo radiation transport codes PHITS and YURI. Monoenergetic organ dose rate coefficients were calculated employing the Monte Carlo code EGSnrc simulating the photon transport in the voxel phantom of a pregnant female and fetus. Photons of initial energies of 0.015–10 MeV were considered including bremsstrahlung. By folding the monoenergetic dose coefficients with the nuclide decay data, nuclide-specific organ doses were obtained. The results of this work can be employed for estimating the doses from external exposures to pregnant women and their fetus, until more precise data are available which include coefficients obtained for phantoms at different stages of pregnancy.
[en] In this paper, translated from an article published by Nuclear Engineering International (NEI), the author presents his views on the future of civil nuclear propulsion: nuclear powered container ships have returned after a decade in the doldrums, nuclear energy already present at sea with more than 200 naval reactors, the development of Floating nuclear power plants, the question of docking of nuclear-powered ships, the possible resistance from incumbent interests at sea in future.
[en] This white paper aims at proposing answers to some questions regarding the situation in Fukushima and the consequences of the accident, notably the situation regarding dismantling, the remaining radioactive substances, lessons learned, and other issues. Thus, a first article presents the radio-ecology discipline and describes how this technique allowed the control and reduction of radionuclide transfers in Fukushima (interview of an IRSN expert). A second article addresses the transport of radioactive substances, more particularly in France where nearly hundred significant events are noticed every year. Classification (in terms of hazard) and regulation (technical requirements and others related to radioactive products and to their transport) aspects are overviewed. The third article addresses the evolution of the dismantling of the Fukushima Daiichi nuclear plant (interview of an IRSN expert). The last article discusses the action plan implemented by France to improve the safety of its nuclear installations, with the installation of back-up Diesel units, the installation of ultimate water sources to cool reactors, and the construction of local crisis centres able to withstand extreme aggressions. Five articles are also proposed. They address the French challenge of management of nuclear wastes, the development of small nuclear reactors everywhere in the world, the necessary adaptation of nuclear plants to extreme air temperatures, the safety of French nuclear installations which is globally good but still to be improved, and the causes and consequences of the Fukushima accident.
[en] Subterranean radioiodine contamination at the Hanford Site in Washington State is believed to be present as iodide, iodate, and organo-I species, with iodate being the predominant form. Because these species have different sediment-sorption characteristics, understanding their distribution is important for developing an accurate understanding of iodine migration in the subsurface. Herein, we report a novel, rapid technique for simultaneous iodine speciation (iodide/iodate) and isotopic ratio (129I/127I) measurements using ion chromatography (IC) joined with collision/reaction cell inductively coupled plasma mass spectrometry (ICP-MS), collectively referred to as IC-ICP-MS. This approach employs online dynamically regenerated eluent suppression post chromatographic separation of the samples and collision cell technology, with pure oxygen as a collision gas for the active suppression of 129Xe (which naturally exists in the argon supplied to the ICP source) to rapidly (< 15 min) achieve precise and reproducible results. Speciated standard reference materials yielded detection limits for 127I of approximately 23.8 ng/L for iodate and 24.3 ng/L for iodide, and for 129I of approximately 1.81 ng/L for iodate and 2.62 ng/L for iodide. The method was demonstrated by analyzing groundwater samples from six wells from 129I-contaminated regions of the Hanford Site; iodate was the primary species for both 127I and 129I. Small quantities of 127I-iodide were also detected in most of the samples, but all 129I-iodide results were below the detection limit. An interference from molybdenum prevented the estimation of organo-iodine concentrations but did not affect the iodate and iodide results. This new analytical capability will enable rapid, simultaneous characterization of speciated inorganic iodine in vadose zone sediments and groundwater samples at levels below the US federal drinking water standard for 129I of 1 pCi/L (∼ 5.6 ng/L). (author)
[en] Described in this paper is an analytic methodology for the solution of the neutron transport equation in slab geometry using PN method. The first part of the present methodology consists of obtaining a local general solution for the PN equations with arbitrary order 𝑵 L ≤ Nand degree 𝑳 ≤ 𝑵 of scattering anisotropy. In the second part, the local general solution for the PN equations was replaced in the scattering source of a simplified version of the linear Boltzmann transport equation, i. e., stationary, slab-geometry, monoenergetic, azimuthally symmetric, for non-multiplying media and isotropic internal source. This methodology has been implemented in a computer code developed on the MatLab® platform for Windows. As a result, in addition to generating numerical results for the scalar flux through the PN method, the computer code generates numerical results for the angular flux at any position in the domain and for any direction not perpendicular to the domain. To evaluate the applicability of the PN method and the analytic methodology, as described in this paper, numerical results for a model problem are presented. (author)
[en] The research was focused on the level and distribution of 90Sr in various parts of the terrestrial environment of Spitsbergen. The mean activity concentrations were noted lower in peats and soils than in cryoconite. Analysis of vertical variation of 90Sr for soils and peats as well as isotopic ratios of 137Cs/90Sr and 239+240Pu/90Sr for cryoconite clearly showed substantial migration or depletion of the considered radionuclide. Due to the large dispersion of isotopic signatures, the 90Sr provenance was difficult to identify in the examined region. However, observed high mobility of the 90Sr might indicate the global fallout origin. (author)
[en] This paper examines radiation-shielding abilities of oxyfluoro-tellurite-zinc glasses in the chemical form of AlF-TeO-ZnO under the substitution of AlF by ZnO. Gamma-ray- and neutron-shielding properties were tested in terms of mass attenuation coefficient (μ/ρ), half value layer, mean free path, effective atomic numbers (Z), effective electron density (N) and removal cross-section (Σ). The μ/ρ values of the glasses were generated by Geant4 simulations over an extended energy range and then the generated data were confirmed via XCOM software. The results showed that both gamma-ray- and neutron-shielding efficiencies of the selected glasses evolved by substituting of AlF by ZnO. Nuclear radiation-shielding abilities of the current glass systems were compared with that of some conventional shielding materials and newly developed HMO glasses. It can be concluded that oxyfluoro-tellurite-zinc glasses could be useful to design novel shields for radiation protection applications.
[en] The hypothesis of the present investigation underlined with determination of possible synergistic effects of serpentine mineral additive on LiBO glasses. A group of LiBO glasses with serpentine mineral additive were synthesized by melt-quenching technique. The elemental analysis of two different LiBO glasses with different amount of serpentine additive is tested using energy-dispersive X-ray (EDX) technique. Next, the surface morphology of synthesized serpentine glasses was investigated with scanning electron microscopy (SEM). The optical features of synthesized serpentine glasses were determined along the wavelength ranged from 200 to 900 nm. Lastly, nuclear radiation shielding properties of LiBO glasses with serpentine mineral additive were determined for gamma rays, neutrons and charged particles. MCNPX (version 2.6.0) general-purpose Monte Carlo code has been utilized for mass attenuation coefficients calculations. The results showed that the spectra are decreasing with wavelength with an observed peak centered at 450 nm. Moreover, it is observed that serpentine mineral additive improves the gamma protecting capacity of LiBO glasses. It was also noticed that the addition of serpentine mineral also enhanced the neutron and charged particle absorption of the glasses.
[en] For 25 LiO–(75 − x) BO–x BiO (where x = 0, 5, 10, 15, 20, 25, 30, 35, and 40 mol%) glasses, gamma-ray and neutrons attenuation features were explored by theoretical approach using ParShield/WinXCOM program, Geant4, and Penelope codes. At Ba (276, 303, 356, and 384 keV), Na (511 and 1280 keV), Cs (662 keV), Mn (835 keV), and Co (1170 and 1330 keV) photon peaks, for all samples, mass attenuation coefficient (μ/ρ), effective atomic number (Z), effective electron density (N), half-value layer (HVL), and mean free path (MFP) parameters have been evaluated using ParShield/WinXCOM program. The μ/ρ values computed by WinXCOM, Geant4, and Penelope codes were compared to check the accuracy, and satisfactory agreement among the values was identified. Moreover, using G–P fitting method as a function of penetration depth (1, 5, 10, 15, 20, 25, 30, 35, and 40 mfp) within the photon energy range of 0.015–15 MeV, exposure buildup factor (EBF) and energy absorption buildup factor (EABF) were derived. For all selected glasses, the effectiveness of the neutrons attenuation has been discussed in terms of macroscopic effective removal cross-section (Σ), coherent scattering cross-section (σ), incoherent scattering cross-section (σ), absorption cross-section (σ), and total neutron cross-section (σ). The 'σ' values have been calculated within 10–10 MeV neutron energy range using the Geant4 code. The μ/ρ possessed larger values at the lowest energy and lower values at higher energy regions for all studied glasses. The μ/ρ, Z, HVL, and MFP values showed enhanced γ-ray shielding capability with BiO content increment in the samples. The 25 LiO–35 BO–40 BiO (mol%) sample by having larger Z and/or Z value, faired lower EBF and EABF values. Largest μ/ρ and Z, and minimal HVL, MFP, EBF, and EABF values of 25 LiO–35 BO–40 BiO (mol%) glass demonstrated its superior γ-ray attenuation ability among all examined glasses. Further, among all glasses, 25 LiO–75 BO (mol%) sample exhibits relatively higher Σ (0.11326 cm) and ‘σ’ (46.109 cm → 0.84607 cm from 1 × 10 MeV → 1×10 MeV neutron energy) values for fast and thermal neutrons attenuation, respectively, indicating its better neutrons absorption competence.
[en] Throughout history, energy has played a fundamental role in human's progress living. To promote nuclear power to meet the future energy needs, ten countries including Argentina, South Africa, the United States, the United Kingdom, Brazil, Japan, Switzerland, France, Canada and Korea in a global effort (Generation IV International Forum - GIF) have agreed to investigate the next generation of nuclear energy systems known as 4 generation. These reactors are expected to enter the market after 2030. Fundamental changes in the configuration of the systems and the forms of the old reactors have led to the production of new reactors, which require basic research and development, careful examination, and the construction of semi-industrial units. The capabilities of fourth-generation reactors are seawater desalination, and thermal applications in addition to the production of electricity. In 2000, the founding countries of GIF formed their first meeting to discuss the need for conduct research on the design of next-generation reactors. Subsequently, a strategy was put forward to direct the activities, and the implementation responsibility was assigned to the US Department of Energy. In this research, we investigate the neutron behavior of the advanced reactor core with lead coolant ALFRED. The purpose of the neutron calculations of the core of a reactor is to calculate the distribution of neutron flux in the center and to calculate the effective reproduction coefficient. Given the necessity of performing lattice pitch neutron calculations, it is initially required to determine the real geometry of the core, as well as the order and fuel richness, the lattice pitch the grid, the radius and height of the fuel rods, the composition and location of the fuel absorbents, the types and locations of the control rods, the fuel complex arrangement, and radial and axial peaking factor. The MCNPX code is used to perform neutron calculations.