970 resultados para Criticality (Nuclear engineering)
Resumo:
A bioprocessing approach for the extraction of base, nuclear and precious metals from refractory and lean grade ores has been reviewed in this paper. Characteristic morphological features of Thiobacillus ferrooxidans, the organism which has been extensively used for biooxidation of sulphide ores have been discussed. Mechanisms of chemoautotrophy and mineral oxidation have been illustrated. The current engineering applications of this microorganism have also been brought out. Various methods for accelerating the growth of Thiobacillus ferrooxidans for faster biooxidation and genetic manipulation for development of desired strains have been outlined.
Resumo:
The present work is an attempt to study crack initiation in nuclear grade, 9Cr-1Mo ferritic steel using AE as an online NDE tool. Laboratory experiments were conducted on 5 heat treated Compact Tension (CT) specimens made out of nuclear grade 9Cr-1Mo ferritic steel by subjecting them to cyclic tensile load. The CT Specimens were of 12.5 mm thickness. The Acoustic emission test system was setup to acquire the data continuously during the test by mounting AE sensor on one of the surfaces of the specimen. This was done to characterize AE data pertaining to crack initiation and then discriminate the samples in terms of their heat treatment processes based on AE data. The AE signatures at crack initiation could conclusively bring to fore the heat treatment distinction on a sample to sample basis in a qualitative sense.Thus, the results obtained through these investigations establish a step forward in utilizing AE technique as an on-line measurement tool for accurate detection and understanding of crack initiation and its profile in 9Cr-1Mo nuclear grade steel subjected to different processes of heat treatment.
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Although some researchers have published friction and wear data of Plasma Nitride (PN) coatings, the tribological behavior of PN/PN Pairs in high vacuum environment has not been published so far In order to bridge this knowledge gap, tribological tests under dry conditions have been conducted on PN/PN Pairs for varying temperatures of 25, 200, 400 and 500 degrees C in high vacuum (1.6 x 10(-4) bar) environment. The PN coatings showed good wear resistance layer on the ring surface. The PN coatings were removed only from the pin surface for all the tests since it contacts at a point. The friction and wear were low at lower temperatures and it eliminated adhesion between the contact surfaces until the coating was completely removed from the pin surface. (C) 2011 Journal of Mechanical Engineering. All rights reserved.
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Ultra-fine crystallites of Mn1-xZnxFe2O4 series (0 <= x <= 1) were synthesized through wet chemical co- precipitation method followed by calcination at 200 degrees C for 4 hours. Formation of ferrites was confirmed by X-ray diffraction, TEM selected area diffraction (SAD) and Fourier Transform Infra-red Spectroscopy (FTIR). Nanocrystallites of different compositions in the series were coated with biocompatible chitosan in order to investigate their possible application as contrast agent for magnetic resonance imaging (MRI). Chitosan coating examined by FTIR, revealed a strong bonding of chitosan molecules to the surface of the ferrite nanocrystallites. Spin-spin, tau(2) relaxivities of nuclear spins of hydrogen protons of the solutions for different ferrites were measured from concentration dependence of relaxation time by nuclear magnetic resonance (NMR). All the compositions of Mn1-xZnxFe2O4 series possess higher values of tau(2) relaxivity thus making them suitable as contrast agents for tau(2) weighted imaging by MRI.
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This paper highlights the seismic microzonation carried out for a nuclear power plant site. Nuclear power plants are considered to be one of the most important and critical structures designed to withstand all natural disasters. Seismic microzonation is a process of demarcating a region into individual areas having different levels of various seismic hazards. This will help in identifying regions having high seismic hazard which is vital for engineering design and land-use planning. The main objective of this paper is to carry out the seismic microzonation of a nuclear power plant site situated in the east coast of South India, based on the spatial distribution of the hazard index value. The hazard index represents the consolidated effect of all major earthquake hazards and hazard influencing parameters. The present work will provide new directions for assessing the seismic hazards of new power plant sites in the country. Major seismic hazards considered for the evaluation of the hazard index are (1) intensity of ground shaking at bedrock, (2) site amplification, (3) liquefaction potential and (4) the predominant frequency of the earthquake motion at the surface. The intensity of ground shaking in terms of peak horizontal acceleration (PHA) was estimated for the study area using both deterministic and probabilistic approaches with logic tree methodology. The site characterization of the study area has been carried out using the multichannel analysis of surface waves test and available borehole data. One-dimensional ground response analysis was carried out at major locations within the study area for evaluating PHA and spectral accelerations at the ground surface. Based on the standard penetration test data, deterministic as well as probabilistic liquefaction hazard analysis has been carried out for the entire study area. Finally, all the major earthquake hazards estimated above, and other significant parameters representing local geology were integrated using the analytic hierarchy process and hazard index map for the study area was prepared. Maps showing the spatial variation of seismic hazards (intensity of ground shaking, liquefaction potential and predominant frequency) and hazard index are presented in this work.
Resumo:
Bentonite clay is identified as potential buffer in deep geological repositories (DGR) that store high level radioactive wastes (HLW) as the expansive clay satisfies the expected mechanical and physicochemical functions of the buffer material. In the deep geological disposal of HLW, iodine-129 is one of the significant nuclides, attributable to its long half-life (half life 1⁄4 1:7 × 107 years). However, the negative charge on the basal surface of bentonite particles precludes retention of iodide anions. To render the bentonite effective in retaining hazardous iodide species in DGR, improvement of the anion retention capacity of bentonite becomes imperative. The iodide retention capac-ity of bentonite is improved by admixing 10 and 20% Ag-kaolinite (Ag-K) with bentonite (B) on a dry mass basis. The present study produced Ag-kaolinite by heating silver nitrate-kaolinite mixes at 400°C. Marginal release of iodide retained by Ag-kaolinite occurred under extreme acidic (pH 1⁄4 2:5) and alkaline (pH 1⁄4 12:5) conditions. The swell pressure and iodide etention results of the B-Ag-K specimens bring out that mixing Ag-K with bentonite does not chemically modify the expansive clay; the mixing is physical in nature and Ag-K presence only contributes to iodide retention of the admixture. DOI: 10.1061/(ASCE)HZ.2153-5515.0000121. © 2012 American Society of Civil Engineers.
Resumo:
One of the main problems of fusion energy is to achieve longer pulse duration by avoiding the premature reaction decay due to plasma instabilities. The control of the plasma inductance arises as an essential tool for the successful operation of tokamak fusion reactors in order to overcome stability issues as well as the new challenges specific to advanced scenarios operation. In this sense, given that advanced tokamaks will suffer from limited power available from noninductive current drive actuators, the transformer primary coil could assist in reducing the power requirements of the noninductive current drive sources needed for current profile control. Therefore, tokamak operation may benefit from advanced control laws beyond the traditionally used PID schemes by reducing instabilities while guaranteeing the tokamak integrity. In this paper, a novel model predictive control (MPC) scheme has been developed and successfully employed to optimize both current and internal inductance of the plasma, which influences the L-H transition timing, the density peaking, and pedestal pressure. Results show that the internal inductance and current profiles can be adequately controlled while maintaining the minimal control action required in tokamak operation.
Resumo:
The growing interest in innovative reactors and advanced fuel cycle designs requires more accurate prediction of various transuranic actinide concentrations during irradiation or following discharge because of their effect on reactivity or spent-fuel emissions, such as gamma and neutron activity and decay heat. In this respect, many of the important actinides originate from the 241Am(n,γ) reaction, which leads to either the ground or the metastable state of 242Am. The branching ratio for this reaction depends on the incident neutron energy and has very large uncertainty in the current evaluated nuclear data files. This study examines the effect of accounting for the energy dependence of the 241Am(n,γ) reaction branching ratio calculated from different evaluated data files for different reactor and fuel types on the reactivity and concentrations of some important actinides. The results of the study confirm that the uncertainty in knowing the 241Am(n,γ) reaction branching ratio has a negligible effect on the characteristics of conventional light water reactor fuel. However, in advanced reactors with large loadings of actinides in general, and 241Am in particular, the branching ratio data calculated from the different data files may lead to significant differences in the prediction of the fuel criticality and isotopic composition. Moreover, it was found that neutron energy spectrum weighting of the branching ratio in each analyzed case is particularly important and may result in up to a factor of 2 difference in the branching ratio value. Currently, most of the neutronic codes have a single branching ratio value in their data libraries, which is sometimes difficult or impossible to update in accordance with the neutron spectrum shape for the analyzed system.
Resumo:
Coupled Monte Carlo depletion systems provide a versatile and an accurate tool for analyzing advanced thermal and fast reactor designs for a variety of fuel compositions and geometries. The main drawback of Monte Carlo-based systems is a long calculation time imposing significant restrictions on the complexity and amount of design-oriented calculations. This paper presents an alternative approach to interfacing the Monte Carlo and depletion modules aimed at addressing this problem. The main idea is to calculate the one-group cross sections for all relevant isotopes required by the depletion module in a separate module external to Monte Carlo calculations. Thus, the Monte Carlo module will produce the criticality and neutron spectrum only, without tallying of the individual isotope reaction rates. The onegroup cross section for all isotopes will be generated in a separate module by collapsing a universal multigroup (MG) cross-section library using the Monte Carlo calculated flux. Here, the term "universal" means that a single MG cross-section set will be applicable for all reactor systems and is independent of reactor characteristics such as a neutron spectrum; fuel composition; and fuel cell, assembly, and core geometries. This approach was originally proposed by Haeck et al. and implemented in the ALEPH code. Implementation of the proposed approach to Monte Carlo burnup interfacing was carried out through the BGCORE system. One-group cross sections generated by the BGCORE system were compared with those tallied directly by the MCNP code. Analysis of this comparison was carried out and led to the conclusion that in order to achieve the accuracy required for a reliable core and fuel cycle analysis, accounting for the background cross section (σ0) in the unresolved resonance energy region is essential. An extension of the one-group cross-section generation model was implemented and tested by tabulating and interpolating by a simplified σ0 model. A significant improvement of the one-group cross-section accuracy was demonstrated.
Resumo:
There is a growing need for very small nuclear reactors for space applications and as portable high-intensity neutron sources. This technical note investigates the question of what is the smallest possible thermal reactor. It was found that the smallest reactor is a spherically shaped solution of 242mAm(NO3)3 in water. The weight of such a reactor is 4.95 kg with 0.7 kg of 242mAm nuclear fuel. The radius of the reactor in this case is 9.6 cm.
Resumo:
There is a growing interest in using 242mAm as a nuclear fuel. The advantages of 242mAm as a nuclear fuel derive from the fact that 242mAm has the highest thermal fission cross section. The thermal capture cross section is relatively low and the number of neutrons per thermal fission is high. These nuclear properties make it possible to obtain nuclear criticality with ultra-thin fuel elements. The possibility of having ultra-thin fuel elements enables the use of these fission products directly, without the necessity of converting their energy to heat, as is done in conventional reactors. There are three options of using such highly energetic and highly ionized fission products. 1. Using the fission products themselves for ionic propulsion. 2. Using the fission products in an MHD generator, in order to obtain electricity directly. 3. Using the fission products to heat a gas up to a high temperature for propulsion purposes. In this work, we are not dealing with a specific reactor design, but only calculating the minimal fuel elements' thickness and the energy of the fission products emerging from these fuel elements. It was found that it is possible to design a nuclear reactor with a fuel element of less than 1 μm of 242mAm. In such a fuel element, 90% of the fission products' energy can escape.