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Hydrogen accommodation in Zr second phase particles: Implications for H pick-up and hydriding of Zircaloy-2 and Zircaloy-4  [PDF]
P. A. Burr,S. T. Murphy,S. C. Lumley,M. R. Wenman,R. W. Grimes
Physics , 2013, DOI: 10.1016/j.corsci.2012.11.036
Abstract: Ab-initio computer simulations have been used to predict the energies associated with the accommodation of H atoms at interstitial sites in {\alpha}, {\beta}-Zr and Zr.M intermetallics formed with common alloying additions (M = Cr, Fe, Ni). Intermetallics that relate to the Zr2(Ni,Fe) second phase particles (SPPs) found in Zircaloy-2 exhibit favourable solution enthalpies for H. The intermetallic phases that relate to the Zr(Cr,Fe)2 SPPs, found predominantly in Zircaloy-4, do not offer favourable sites for interstitial H. It is proposed that Zr(Cr,Fe)2 particles may act as bridges for the migration of H through the oxide layer, whilst the Zr2(Ni,Fe)-type particles will trap the migrating H until these are dissolved or fully oxidised.
Computation of the Optimal Value of Operating Parameters in a Reactor – Heat Exchanger System by Differential Evolution Techniques  [PDF]
Gopalakrishnan.B,Dr.P.K.Bhaba
International Journal of Recent Technology and Engineering , 2013,
Abstract: In this research work, the modern soft computingtechnique of Differential Evolution (DE) algorithm is consideredto determine global optimal values of the operating parameters ina Reactor -Heat Exchanger (RHE) system. In addition, a penaltyterm is incorporated in the objective function and therebycomputing annual cost of the RHE system in terms of operatingand investment costs. A comparative study is also made withGenetic Algorithm (GA) in RHE system. Results clearly indicatethe supremacy of DE for global optimization of operatingparameters in RHE system. A convergence test is performed andreported here.
Tensile Behavior of Cryorolled Zircaloy-2
American Journal of Materials Science , 2012, DOI: 10.5923/j.materials.20120205.01
Abstract: Zircaloy-2 is mainly used in nuclear technology, as cladding of fuel rods in nuclear reactors, especially water reactors (BWRs). Hence high strength of Zircaloy-2 is of prime importance. This investigation deals with the effect of cryorolling on Zircaloy-2 by comparing different tensile properties. For this analysis, four samples with various degrees of cryorolling are taken and tensile tests are conducted on these samples. The obtained results are analyzed and the optimum degree of cryorolling of Zircaloy-2 is obtained. The cryorolling improved the mechanical properties of the material as the dislocations are entangled near the grain boundaries and also due to decrease in the grain size. The microstructure of the sample is analysed by optical microscope, before and after cryorolling and the grain structure analysis is done.
Optimization of Sequencing Batch Reactor Operating Conditions for Treatment of High-strength Pharmaceutical Wastewater  [PDF]
Emad S. Elmolla,Natasha Ramdass,Malay Chaudhuri
Journal of Environmental Science and Technology , 2012,
Abstract: Optimization of the Sequencing Batch Reactor (SBR) operating conditions for high-strength non-penicillin pharmaceutical wastewater treatment is a challenging task. Two SBR were operated under different operating conditions. Three different Hydraulic Retention Times (HRT) (12, 24 and 48 h) were tested and operated under high (6000 mg L-1) and low (4000 mg L-1) Mixed Liquor Suspended Solids (MLSS) concentration. Statistical analysis (two-way ANOVA) using SPSS software was applied on the results to evaluate the effect of HRT and MLSS concentration on the SBR performance. The hydraulic retention time of 24 h was found suitable for the SBR and increasing HRT from 12 to 24 h significantly affected the BOD5 removal; however, increasing HRT from 24 to 48 h or increasing MLSS concentration from 4000 to 6000 mg L-1 did not significantly improve the SBR performance in terms of BOD5 removal. Under the optimum operating conditions (HRT 24 h and MLSS 4000 mg L-1), the SBR achieved an efficiency of 94±1.3% in terms of BOD5 removal and 83.9±1.7% in terms of COD removal, with complete nitrification.
Some questions on nuclear safety of heavy-water power reactor operating in self-sufficient thorium cycle  [PDF]
Bergelson Boris R.,Gerasimov Alexander S.,Tikhomirov Georgy V.
Nuclear Technology and Radiation Protection , 2008, DOI: 10.2298/ntrp0802022b
Abstract: In this paper the comparative calculations of the void coefficient have been made for different types of channel reactors for the coolant density interval 0.8-0.01 g/cm3. These results demonstrate the following. In heavy-water channel reactors, the replacement of D2O coolant by H2O, ensuring significant economic advantage, leads to the essential reducing of nuclear safety of an installation. The comparison of different reactors by the void coefficient demonstrates that at the dehydration of channels the reactivity increase is minimal for HWPR(Th), operating in the self-sufficient mode. The reduction of coolant density in channels in most cases is accompanied by the increase of power and temperatures of fuel assemblies. The calculations show that the reduction of reactivity due to Doppler effect can compensate the effect of dehydration of a channel. However, the result depends on the time dependency of heat-hydraulic processes, occurring in reactor channels in the specific accident. The result obtained in the paper confirms that nuclear safety of HWPR(Th) lies on the same level as nuclear safety of CANDU type reactors approved in practice.
A Simulation Study of the Steam Reforming of Methaneina Fixed-Bed Reactor  [PDF]
Fernando Ant?niode Araújo Silva, Kenia Carvalho Mendes, Jornandes Dias da Silva
Engineering (ENG) , 2016, DOI: 10.4236/eng.2016.84021
Abstract: In this work a one-dimensional mathematical model was developed to simulate methane conversion and hydrogen yield in a fixed-bed reactor filled with catalyst particles. For the reason that reforming reactions are sorely endothermic process, the heat is supplied to the reactor through electrical heating. The reforming reactions have been investigated from a modelling view point considering the effect of different temperatures ranging from 500 and 977 on the conversion of methane and hydrogen yield. Simulation results show that the steam reforming of methane in a fixed-bed reactor can efficiently store high temperature end thermal energy. When the operating temperature is increased to 977, the conversion of methane is 97.48% and the hydrogen yield is 2.2408. As a conclusion, the maximum thermochemical efficiency will be obtained under optimal operating temperature (977) and the steam/methane (3.86) ratio.
The Significance of Carbon 14 in Graphite Reactor Components at End of Generation  [PDF]
Martin Metcalfe, Athanasia Tzelepi
Journal of Environmental Protection (JEP) , 2019, DOI: 10.4236/jep.2019.102008
Abstract: It is estimated that there are at least 250,000 tonnes of irradiated graphite worldwide that will require eventual disposal. This graphite arises from a number of sources, but principally comprises moderator and structural materials for experimental reactors, production reactors, commercial power reactors and fuel assemblies. In the UK, a significant proportion of its irradiated graphite is classified as Intermediate Level Waste. Such waste is not heat generating but has a radioactive content exceeding 4 GBq per tonne alpha or 12 GBq per tonne beta/gamma activity. While the classification of waste is not consistent across states and proposals by individual states for the management of their graphite waste vary considerably, a common interest is the nature and distribution of its radioactive content. The radionuclides in irradiated graphite presenting the most significant long-term hazard are Carbon 14 (C-14) and Chlorine 36 (Cl-36) with half-lives of 5730 and 301,000 years respectively. For a better understanding of the way in which C-14 is produced, its distribution within irradiated graphite and realistic quantification of activity can potentially lead to improved characterization to validate its status within current or future waste classifications, segregation to reduce Intermediate Level Waste volumes, or treatment to reduce activity enabling re-classification as Low Level Waste. This paper reviews all these issues and then focuses on the significance of C-14. Some findings from a National Nuclear Laboratory study of C-14 levels in carbonaceous deposits and the underlying Magnox reactor graphite are presented to illustrate the need for thorough characterization of the waste material. These results are discussed in the context of aqueous leaching of C-14 from irradiated graphite and potential treatment options to minimize aqueous release. The paper concludes with some broader observations on the significance of C-14 in nuclear reactor graphite components and how these issues should be considered when preparing the lifetime management of new nuclear plant.
Electrochemical and Oxidation Behavior of Yttria Stabilized Zirconia Coating on Zircaloy-4 Synthesized via Sol-Gel Process  [PDF]
S. Rezaee,Gh. R. Rashed,M. A. Golozar
International Journal of Corrosion , 2013, DOI: 10.1155/2013/453835
Abstract: Sol-gel 8?wt.% Yttria Stabilized Zirconia (YSZ) thin films were prepared on zirconium (zircaloy-4?alloy) by dip-coating technique followed by heat treating at various temperatures (200°C, 400°C, and 700°C) in order to improve both electrochemical corrosion and high temperature oxidation properties of the substrate. Differential thermal analysis and thermogravimetric analysis (DTA-TG) revealed the coating formation process. X-ray diffraction (XRD) was used to determine the crystalline phase structure transformation. The morphological characterization of the coatings was carried out using scanning electron microscopy (SEM). The electrochemical behavior of the coated and uncoated samples was investigated by means of open circuit potential, Tafel, and electrochemical impedance spectroscopy (EIS) in a 3.5?wt.% NaCl solution. The homogeneity and surface appearance of coatings produced was affected by the heat treatment temperature. According to the corrosion parameters, the YSZ coatings showed a considerable increase in the corrosion resistance, especially at higher heat treatment temperatures. The coating with the best quality, from the surface and corrosion point of view, was subjected to oxidation test in air at 800°C. The coated sample presented a 25% reduction in oxidation rate in comparison with bare substrate. 1. Introduction Because of many excellent bulk properties like low thermal neutron capture cross section, favorable mechanical properties and good corrosion resistance even at high temperatures, zirconium and its alloys are widely used in nuclear reactors as fuel cladding and as reactor structural elements, chemical engineering, and lately in biomedical applications. For fuel cladding applications, the alloys of choice are zircaloy-2 (Zr-2) and zircaloy-4 (Zr-4). In fuel deep geological repository, used fuel bundles and the associated Zr cladding are encapsulated in durable containers and the containers are sealed in an engineered vault at a depth of hundreds of meters in a stable low permeability rock mass. Ground water composition especially in crystalline and sedimentary rock types contains NaCl. In a failed container, zirconium is in contact with NaCl solution [1–4]. So, they are somehow subject to corrosion and their specific surface properties (e.g., corrosion, oxidation, etc.) should be improved [2, 4]. Surface modification of materials permits independent optimization of bulk and surface properties. Among all surface modification techniques such as chemical vapor deposition [5, 6], physical vapor deposition [7], plasma [8], ion
Specific Aspects of Internal Corrosion of Nuclear Clad Made of Zircaloy  [PDF]
Jean-Baptiste Minne,Lionel Desgranges,Virgil Optasanu,Nathalie Largenton,Laura Raceanu,Tony Montesin
Physics , 2013, DOI: 10.4028/www.scientific.net/DDF.323-325.227
Abstract: In PWR, the Zircaloy based clad is the first safety barrier of the fuel rod, it must prevent the dispersion of the radioactive elements, which are formed by fission inside the UO2 pellets filling the clad. We focus here on internal corrosion that occurs when the clad is in tight contact with the UO2 pellet. In this situation, with temperature of 400^{\circ}C on the internal surface of the clad, a layer of oxidised Zircaloy is formed with a thickness ranging from 5 to 15 $\mu$m. In this paper, we will underline the specific behaviour of this internal corrosion layer compared to wet corrosion of Zircaloy. Simulations will underline the differences of stress field and their influences on corresponding dissolved oxygen profiles. The reasons for these differences will be discussed as function of the mechanical state at inner surface of the clad which is highly compressed. Differences between mechanical conditions generated by an inner or outer corrosion of the clad are studied and their influences on the diffusion phenomena are highlighted.
Pick interpolation in several variables  [PDF]
Ryan Hamilton
Mathematics , 2011,
Abstract: We investigate the Pick problem for the polydisk and unit ball using dual algebra techniques. Some factorization results for Bergman spaces are used to describe a Pick theorem for any bounded region in $\mathbb{C}^d$.
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