Relevant bibliographies by topics / Core parameters

Academic literature on the topic 'Core parameters'

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Published: 10 December 2022
Last updated: 28 January 2023

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Journal articles on the topic "Core parameters"

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Lothe, J., and J. P. Hirth. "Dislocation core parameters." physica status solidi (b) 242, no. 4 (March 2005): 836–41. http://dx.doi.org/10.1002/pssb.200402114.

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Baumgardt, Holger, Douglas C. Heggie, Piet Hut, and Junichiro Makino. "Parameters of core collapse." Monthly Notices of the Royal Astronomical Society 341, no. 1 (May 2003): 247–50. http://dx.doi.org/10.1046/j.1365-8711.2003.06407.x.

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Schoeck, Gunther. "Atomic dislocation core parameters." physica status solidi (b) 247, no. 2 (December 10, 2009): 265–68. http://dx.doi.org/10.1002/pssb.200945379.

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Luthfi, Wahid, and Surian Pinem. "VALIDATION OF SRAC CODE SYSTEM FOR NEUTRONIC PARAMETERS CALCULATION OF THE PWR MOX/UO2 CORE BENCHMARK." Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir 27, no. 1 (February 28, 2021): 47. http://dx.doi.org/10.17146/urania.2021.27.1.6238.

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VALIDATION OF SRAC CODE SYSTEM FOR NEUTRONIC PARAMETERS CALCULATION OF THE PWR MOX/UO2 CORE BENCHMARK. Determination of neutronic parameter value is an important part in determining reactor safety, so accurate calculation results can be obtained. This study is focused on the validation of SRAC code system in the calculation of neutronic parameters value of a PWR (Pressurized Water Reactor) reactor core. MOX/UO2 Core Benchmark was choosed because it is used by several researchers as a reference core for code validation in the determination of neutronic parameters of a reactor core. The neutronic parameters calculated include critical boron concentration, delayed neutron fraction, and Power Peaking Factor (PPF) and its distribution in axial and radial directions. When compared with reference data, the calculation results of the critical boron concentration value show that there is a difference of 22.5 ppm on SRAC code system. Meanwhile, differences in power per fuel element (assembly power error) value of power-weighted error (PWE) and error-weighted error (EWE) is 2.93% and 3.94%, respectively. Maximum difference between PPF value in axial direction with reference reaches a value of 4.57%. SRAC calculation results also show consistency with the calculation results of other program packages or code. Results of this study indicate that SRAC code system is still quite accurate for the calculation of neutronic parameters of PWR reactor core benchmark. Therefore, SRAC code system can be used to calculate neutronic parameters of PWR reactor core, especially when using MOX (mixed oxide) fuel.Keywords: Neutronic parameter, critical boron concentration, power peaking factor, SRAC code system.
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Mitkov, Svetlomir, Ivan Spasov, and Nikola Kolev. "Thermal-hydraulic analysis of a VVER-1000 core in MSLB conditions." E3S Web of Conferences 327 (2021): 01013. http://dx.doi.org/10.1051/e3sconf/202132701013.

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The objective of this paper is to analyze the ability of a VVER-1000 core and its control system to cope with a hypothetical main steam line break (MSLB) accident in case of multiple equipment failures. The study involves the use of advanced 3D core calculation models benchmarked and validated for reactivity accidents in preceding studies. A MSLB core boundary condition problem is solved on a coarse (nodal) mesh with the coupled COBAYA/CTF neutronic/thermal hydraulic codes. The core thermal-hydraulic boundary conditions are obtained from a preceding full-plant MSLB simulation. The assessment of the core safety parameters is supplemented by a fine-mesh (sub-channel) thermal-hydraulic analysis of the hottest assemblies with the CTF code using information from the 3D nodal COBAYA/CTF calculations. Thirteen variants of a pessimistic MSLB scenario are considered, each of them assuming a number of equipment failures aggravated by eight control rods stuck out of the core after scram at different locations in the overcooled sector. The results (within the limitations of the adopted modeling assumptions) show that the core safety parameters do not exceed the safety limits in the simulated aggravated reactivity accidents.
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Valentino, E. Di, T. Brinckmann, M. Gerbino, V. Poulin, F. R. Bouchet, J. Lesgourgues, A. Melchiorri, et al. "Exploring cosmic origins with CORE: Cosmological parameters." Journal of Cosmology and Astroparticle Physics 2018, no. 04 (April 5, 2018): 017. http://dx.doi.org/10.1088/1475-7516/2018/04/017.

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Novak, Franc, Peter Mrak, and Anton Biasizzo. "Measuring Static Parameters of Embedded ADC Core." Journal of Electrical Engineering 62, no. 2 (March 1, 2011): 80–86. http://dx.doi.org/10.2478/v10187-011-0013-3.

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Measuring Static Parameters of Embedded ADC CoreThe paper presents the results of a feasibility study of measuring static parameters of ADC cores embedded in a System-on-Chip. Histogram based technique is employed because it is suitable for built-in self-test. While the theoretical background of the technique has been covered by numerous papers, less attention has been given to implementations in practice. Our goal was the implementation of histogram test in a IEEE Std 1500 wrapper. Two different solutions pursuing either minimal test time or minimal hardware overhead are described. The impact of MOS switches at ADC input on the performed measurements was considered.
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Zografos, George C., Flora Zagouri, Theodoros N. Sergentanis, Afrodite Nonni, Nikolaos Lymperopoulos, and Effstratios Patsouris. "What parameters affect pain in core biopsy?" European Radiology 18, no. 6 (February 16, 2008): 1144–45. http://dx.doi.org/10.1007/s00330-008-0879-z.

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Turinsky, Paul, Aaron Graham, Hisham Sarsour, and Benjamin Collins. "GENERATION OF NODAL CORE SIMULATOR UTILIZING VERA." EPJ Web of Conferences 247 (2021): 02018. http://dx.doi.org/10.1051/epjconf/202124702018.

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Nuclear core simulators based upon few-group nodal diffusion method currently are widely used to predict light water reactor core behavior. Nodal parameters’ input, e.g. cross-sections, discontinuity factors, and pin form factors, are typically generated utilizing lattice physics codes. In doing so, a number of approximations are introduced related to using zero current boundary conditions, 2-D radial geometry, and uniform thermal conditions in coolant and fuel. Usage of full core models with prediction fidelity typical of lattice physics to predict nodal parameters would eliminate these approximations. The VERA code can serve as such a full core model and was so utilized in this work. Via subsequent processing of VERA predictions, for a range of state points, nodal parameters and their functionalization in terms of coolant density, fuel temperature, and soluble poison concentration, were obtained and input to the NESTLE nodal code. By processing VERA predictions using consistent nodal methodologies as used in NESTLE, when using nodal parameters after functionalization based upon All-Rods-Out (ARO) VERA state points, the maximum reactivity and pin power differences between VERA and NESTLE were 2 pcm and 0.003 for ARO core simulations. For rodded core simulations, these maximum differences grew to 58 pcm and 0.04. Increases in differences were determined to be attributed to usage of unrodded nodal parameters generated using the VERA ARO state points whether the core is partially rodded or not, consistent with lattice physics practice. Obtaining unrodded nodal parameters using the VERA rodded state points reduced maximum differences to 2 pcm and 0.003 in pin powers.
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Ashoub, N., and E. Amin. "Neutronic studies for the second Egyptian research reactor ET-RR-2." Kerntechnik 66, no. 4 (August 1, 2001): 177–81. http://dx.doi.org/10.1515/kern-2001-0077.

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Abstract The start up physics studies SPS of the Second Egyptian Research Reactor is presented in this work. The SPS consists of the approach to criticality, the power and flux distribution of the first critical core, criticality, power and flux distribution of the working core. The working core is the beryllium-reflected core with cobalt loaded for irradiation. Some of the neutronic parameters are determined, these parameters are the power peaking factor, power distribution, the excess reactivity for both the water and beryllium reflector cores, the core excess reactivity with the cobalt loaded, the reactivity worth of the cobalt irradiation box. Flux distribution for one fast, three eipthermal and one thermal energy group are also determined. All the above neutronic parameters are calculated by using the lattice cell code WIMSD4 and the two dimensional diffusion code DIXY2.
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Dissertations / Theses on the topic "Core parameters"

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Kersting, Alyssa (Alyssa Rae). "Fluence-limited burnup as a function of fast reactor core parameters." Thesis, Massachusetts Institute of Technology, 2011. http://hdl.handle.net/1721.1/76938.

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Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2011.
Cataloged from PDF version of thesis.
Includes bibliographical references (p. 39).
The limiting factor in current designs for fast reactors is not only the reactivity, but also the maximum permissible fast-neutron fluence in the cladding, especially for reduced uranium enrichment cores using high-albedo MgO reflectors. The intent of this thesis was to determine the best design parameters - fuel type, fuel compound, fuel arrangement, and coolant - while observing these limitations. The ERANOS code was used to determine the flux values for each design option. A curve was fitted to the fluxes taken at beginning of life, middle of life, and end of life. This curve was then integrated progressively until the clad fluence limit of 4 x 1023 fast neutrons/cm 2 was reached. The different design options were compared with emphasis on minimizing enrichment and maximizing burnup. Sodium was chosen as a coolant because of its extensive experience based compared to the other options, as well as its heat transfer properties. Inverted fuel was found to be better neutronically, in both clad lifetime and burnup than conventional pin-type fuel, but the requirement of fuel venting may discourage use of this option. Uranium carbide was found to be superior to nitride, oxide or metal fuel in its clad lifetime, especially if pin cell fuel is used. If inverted fuel is used, uranium oxide is also a valid choice from a burnup and cost perspective, especially should re-cladding not be feasible or desired, since the reactivity and clad fluence lifetimes of oxide fuel are similar to each other.
by Alyssa Kersting.
S.B.
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Moloko, Lesego Ernest. "Impact of beryllium reflector ageing on Safari–1 reactor core parameters / L.E. Moloko." Thesis, North-West University, 2011. http://hdl.handle.net/10394/6930.

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The build–up of 6Li and 3He, that is, the strong thermal neutron absorbers or the so called "neutron poisons", in the beryllium reflector changes the physical characteristics of the reactor, such as reactivity, neutron spectra, neutron flux level, power distribution, etc.; furthermore,gaseous isotopes such as 3H and 4He induce swelling and embrittlement of the reflector. The SAFARI–1 research reactor, operated by Necsa at Pelindaba in South Africa, uses a beryllium reflector on three sides of the core, consisting of 19 beryllium reflector elements in total. This MTR went critical in 1965, and the original beryllium reflectors are still used. The individual neutron irradiation history of each beryllium reflector element, as well as the impact of beryllium poisoning on reactor parameters, were never well known nor investigated before. Furthermore, in the OSCAR{3 code system used in predictive neutronic calculations for SAFARI–1, beryllium reflector burn–up is not accounted for; OSCAR models the beryllium reflector as a non–burnable, 100% pure material. As a result, the poisoning phenomenon is not accounted for. Furthermore, the criteria and hence the optimum replacement time of the reflector has never been developed. This study presents detailed calculations, using MCNP, FISPACT and the OSCAR{3 code system, to quantify the influence of impurities that were originally present in the fresh beryllium reflector, the beryllium reflector poisoning phenomenon, and further goes on to propose the reflector's replacement criteria based on the calculated fluence and predicted swelling. Comparisons to experimental low power flux measurements and effects of safety parameters are also established. The study concludes that, to improve the accuracy and reliability of the predictive OSCAR code calculations, beryllium re flector burn–up should undoubtedly be incorporated in the next releases of OSCAR. Based on this study, the inclusion of the beryllium reflector burn–up chains is planned for implementation in the currently tested OSCAR–4 code system. In addition to beryllium reflector poisoning, the replacement criteria of the reflector is developed. It is however crucial that experimental measurements on the contents of 3H and 4He be conducted and thus swelling of the reflector be quantifed. In this way the calculated results could be verified and a sound replacement criteria be developed. In the absence of experimental measurements on the beryllium reflector, the analysis and quantifcation of the calculated results is reserved for future studies.
Thesis (M.Sc. Engineering Sciences (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2011.
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Kalathil, Robins T. "Assessment of Uncertainty in Core Body Temperature due to Variability in Tissue Parameters." University of Cincinnati / OhioLINK, 2016. http://rave.ohiolink.edu/etdc/view?acc_num=ucin1479819529740889.

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Zhang, Rui. "Transformer modelling and influential parameters identification for geomagnetic disturbances events." Thesis, University of Manchester, 2012. https://www.research.manchester.ac.uk/portal/en/theses/transformer-modelling-and-influential-parameters-identification-for-geomagnetic-disturbances-events(e7c8df5c-8fa9-491f-bc06-9cb90cbbf543).html.

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Power transformers are a key element in the transmission and distribution of electrical energy and as such need to be highly reliable and efficient. In power system networks, transformer core saturation can cause system voltage disturbances or transformer damage or accelerate insulation ageing. Low frequency switching transients such as ferroresonance and inrush currents, and increasingly what is now known as geomagnetic induce currents (GIC), are the most common phenomena to cause transformer core saturation. This thesis describes extensive simulation studies carried out on GIC and switching ferroresonant transient phenomena. Two types of transformer model were developed to study core saturation problems; one is the mathematical transformer magnetic circuit model, and the other the ATPDraw transformer model. Using the mathematical transformer magnetic circuit model, the influence of the transformer core structure on the magnetising current has been successfully identified and so have the transformers' responses to GIC events. By using the ATPDraw transformer model, the AC system network behaviours under the influence of the DC bias caused by GIC events have been successfully analysed using various simulation case studies. The effects of the winding connection, the core structure, and the network parameters including system impedances and transformer loading conditions on the magnetising currents of the transformers are summarised. Transient interaction among transformers and other system components during energisation and de-energisation operations are becoming increasingly important. One case study on switching ferroresonant transients was modelled using the available transformer test report data and the design data of the main components of the distribution network. The results were closely matched with field test results, which verified the simulation methodology. The simulation results helped establish the fundamental understanding of GIC and ferroresonance events in the power networks; among all the influential parameters identified, transformer core structure is the most important one. In summary, the five-limb core is easier to saturate than the three-limb transformer under the same GIC events; the smaller the side yoke area of the five-limb core, the easier it will be to saturate. More importantly, under GIC events a transformer core could become saturated irrespective of the loading condition of the transformer.
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Polzer, Gudrun [Verfasser]. "Analyse von Erdgezeitenbeobachtungen zur Bestimmung der Parameter der Erdkernresonanz = Analysis of earth tide observations for the determination of core resonance parameters / Gudrun Polzer." Karlsruhe : KIT-Bibliothek, 1997. http://d-nb.info/1013806271/34.

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Sbiga, Hassan M. "Prediction and measurement of special core analysis petrophysical parameters in the Nubian sandstone of the North Africa." Thesis, Heriot-Watt University, 2013. http://hdl.handle.net/10399/2677.

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One of the main objectives of this work was to investigate the applicability and accuracy of artificial neural networks for estimating special core analysis (SCAL) parameters from minimal core training data and wireline logs. The SCAL data was obtained from measurements on core plugs undertaken at the Libyan Petroleum Institute (L.P.I). Previous neural network studies have attempted to predict routine core analysis parameters, such as permeability, but not SCAL parameters such as true formation resistivity (Rt), resistivity index (RI), water saturation (Sw), saturation exponent (n) and Amott-Harvey Wettability Index (IA/H). Different combinations of wireline logs were used to train a variety of neural network predictors. Some of the predictors were trained using a large dataset from the entire cored interval of the training well. Other genetically focused neural network (GFNN) predictors were trained just from one short representative genetic unit (RGU) in the training well. The predictors were then tested in an adjacent well in the same oil field and also in another well in a different oil field. Significantly the performance of the GFNN predictors was as good (and in most cases better) than the predictors trained on the much larger dataset. This demonstrated the useful of the GFNN approach, which is very cost effective in terms of the minimal core that is required, and the reduced computer processing time. Moreover, this is the first time that these GFNN predictors have been used to predict SCAL parameters in the studied area, the Nubian Sandstone Formation in North Africa. These neural network predictors are particularly useful in this area due to the limited amount of SCAL data that is currently available. Quantitative statistical measures of heterogeneity were also examined on the reservoir samples, followed by a comparative analysis of hydraulic units (HUs) with a newer approach of global hydraulic elements (GHEs) to characterize the reservoir units in the studied area. The GHEs were then applied to select minimal representative core training data to train the genetically focused neural networks (GFNNs) to predict the SCAL parameters. The thesis also describes the factors affecting SCAL resistivity parameters. Laboratory measurements on the Nubian Sandstone reservoir rock samples showed changes in the formation resistivity factor (F) and cementation exponent (m) between ambient conditions and at overburden pressures. Changes were also observed in the saturation exponent (n) before and after wettability measurement. The experimental results also showed that there was a good relation between resistivity and the type of pore system which is consistent with study result from Swanson (1985) confirming earlier work.
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Casasola, Raffaella. "Electrospinning of poly (lactic) acid for biomedical applications : analysis of solution properties and process parameters, drug encapsulation and release." Thesis, Loughborough University, 2016. https://dspace.lboro.ac.uk/2134/22949.

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Electrospinning or electrostatic fibre spinning employs electrostatic force to draw fibres from a liquid, either a polymeric solution or a polymer melt in the form of a charged jet. The jet solidifies and is deposited on a collector in the form of a non-woven fibrous mat. Electrospun fibres have diameters between several nanometres to a few microns, which is one of the main advantages of the process, as materials at the nanoscale have shown great potential in a wide range of healthcare and energy applications. The initial selection of solvents to dissolve the polymer for production of electrospun defect-free nanofibres has generally been based on experience from similar polymer-solvent systems. The selection of a solvent is important to control the fibre surface morphology that would eventually determine the field of application for the electrospun nanofibrous structures. However, little attempt has been made to study the correlation between the solubility behaviour of the polymer in different solvents and the electrospinnability of the polymer solutions. From this perspective, the first part of this thesis focused on the selection of different solvents for the production of poly (lactic acid) (PLA) nanofibres. Solution properties were measured and the electrospun nanofibrous structures were analysed in terms of morphology and nanofibre diameter. Understanding the molecular interactions between polymer and solvents enables the correct solvent selection to ensure the desired nanofibrous structure. Solubility is not the only criterion for nanofibre formation from a polymer solution. Polymer concentration is one of the main factors affecting electrospinning. For this reason, a relationship between PLA concentration and nanofibre morphology was determined by solution property measurements. A three step systematic methodology has been proposed in this thesis in order to select appropriate solvent and polymer concentration for the production of homogeneous electrospun mats made of defect-free nanofibres. This methodology was validated for PLA nanofibres, but it can be used for a wide range of polymers. It simplifies the solvent selection process and can significantly improve the trial and error approaches used at present. Despite several models for electrospinning having been proposed to predict the behaviour of the electrospun jet, there are no simple methods for a priori prediction of the final morphology of the electrospun mat from the knowledge of solution properties and electrospinning process parameters. Moreover the prediction of nanofibre diameter is still a difficulty and little research has been done on this. For these reasons, the second part of this thesis is dedicated to understanding the effect of some process parameters on the jet electric current and hence on the morphology of PLA nanofibres. The values of current measured were used to verify an equation proposed in the literature for the prediction of the final diameter. The experimental diameter of the PLA nanofibres was compared with the predicted value. In the last chapter coaxial electrospinning was employed to produce PLA nanofibres with a core shell structure for the incorporation of a model hydrophilic drug in the nanofibre core. The large surface area to volume ratio of nanofibres offers the great advantage of higher efficiency of encapsulation and better control of the release profile compared with other drug delivery systems. Even though successful encapsulation of drug and proteins have been reported, it is not clear how to verify the continuous drug distribution in the core throughout the whole length of the fibre. The solution properties of both core and shell strongly affect the outcome of the electrospinning process. For this reason, several techniques have been used to verify the formation of a core shell structure and proper encapsulation of the drug. In addition, the efficiency of drug encapsulation was evaluated and drug release studies were performed.
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Mesado, Melia Carles. "Uncertainty Quantification and Sensitivity Analysis for Cross Sections and Thermohydraulic Parameters in Lattice and Core Physics Codes. Methodology for Cross Section Library Generation and Application to PWR and BWR." Doctoral thesis, Universitat Politècnica de València, 2017. http://hdl.handle.net/10251/86167.

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This PhD study, developed at Universitat Politècnica de València (UPV), aims to cover the first phase of the benchmark released by the expert group on Uncertainty Analysis in Modeling (UAM-LWR). The main contribution to the benchmark, made by the thesis' author, is the development of a MATLAB program requested by the benchmark organizers. This is used to generate neutronic libraries to distribute among the benchmark participants. The UAM benchmark pretends to determine the uncertainty introduced by coupled multi-physics and multi-scale LWR analysis codes. The benchmark is subdivided into three phases: 1. Neutronic phase: obtain collapsed and homogenized problem-dependent cross sections and criticality analyses. 2. Core phase: standalone thermohydraulic and neutronic codes. 3. System phase: coupled thermohydraulic and neutronic code. In this thesis the objectives of the first phase are covered. Specifically, a methodology is developed to propagate the uncertainty of cross sections and other neutronic parameters through a lattice physics code and core simulator. An Uncertainty and Sensitivity (U&S) analysis is performed over the cross sections contained in the ENDF/B-VII nuclear library. Their uncertainty is propagated through the lattice physics code SCALE6.2.1, including the collapse and homogenization phase, up to the generation of problem-dependent neutronic libraries. Afterward, the uncertainty contained in these libraries can be further propagated through a core simulator, in this study PARCSv3.2. The module SAMPLER -available in the latest release of SCALE- and DAKOTA 6.3 statistical tool are used for the U&S analysis. As a part of this process, a methodology to obtain neutronic libraries in NEMTAB format -to be used in a core simulator- is also developed. A code-to-code comparison with CASMO-4 is used as a verification. The whole methodology is tested using a Boiling Water Reactor (BWR) reactor type. Nevertheless, there is not any concern or limitation regarding its use in any other type of nuclear reactor. The Gesellschaft für Anlagen und Reaktorsicherheit (GRS) stochastic methodology for uncertainty quantification is used. This methodology makes use of the high-fidelity model and nonparametric sampling to propagate the uncertainty. As a result, the number of samples (determined using the revised Wilks' formula) does not depend on the number of input parameters but only on the desired confidence and uncertainty of output parameters. Moreover, the output Probability Distribution Functions (PDFs) are not subject to normality. The main disadvantage is that each input parameter must have a pre-defined PDF. If possible, input PDFs are defined using information found in the related literature. Otherwise, the uncertainty definition is based on expert judgment. A second scenario is used to propagate the uncertainty of different thermohydraulic parameters through the coupled code TRACE5.0p3/PARCSv3.0. In this case, a PWR reactor type is used and a transient control rod drop occurrence is simulated. As a new feature, the core is modeled chan-by-chan following a fully 3D discretization. No other study is found using a detailed 3D core. This U&S analysis also makes use of the GRS methodology and DAKOTA 6.3.
Este trabajo de doctorado, desarrollado en la Universitat Politècnica de València (UPV), tiene como objetivo cubrir la primera fase del benchmark presentado por el grupo de expertos Uncertainty Analysis in Modeling (UAM-LWR). La principal contribución al benchmark, por parte del autor de esta tesis, es el desarrollo de un programa de MATLAB solicitado por los organizadores del benchmark, el cual se usa para generar librerías neutrónicas a distribuir entre los participantes del benchmark. El benchmark del UAM pretende determinar la incertidumbre introducida por los códigos multifísicos y multiescala acoplados de análisis de reactores de agua ligera. El citado benchmark se divide en tres fases: 1. Fase neutrónica: obtener los parámetros neutrónicos y secciones eficaces del problema específico colapsados y homogenizados, además del análisis de criticidad. 2. Fase de núcleo: análisis termo-hidráulico y neutrónico por separado. 3. Fase de sistema: análisis termo-hidráulico y neutrónico acoplados. En esta tesis se completan los principales objetivos de la primera fase. Concretamente, se desarrolla una metodología para propagar la incertidumbre de secciones eficaces y otros parámetros neutrónicos a través de un código lattice y un simulador de núcleo. Se lleva a cabo un análisis de incertidumbre y sensibilidad para las secciones eficaces contenidas en la librería neutrónica ENDF/B-VII. Su incertidumbre se propaga a través del código lattice SCALE6.2.1, incluyendo las fases de colapsación y homogenización, hasta llegar a la generación de una librería neutrónica específica del problema. Luego, la incertidumbre contenida en dicha librería puede continuar propagándose a través de un simulador de núcleo, para este estudio PARCSv3.2. Para el análisis de incertidumbre y sensibilidad se ha usado el módulo SAMPLER -disponible en la última versión de SCALE- y la herramienta estadística DAKOTA 6.3. Como parte de este proceso, también se ha desarrollado una metodología para obtener librerías neutrónicas en formato NEMTAB para ser usadas en simuladores de núcleo. Se ha realizado una comparación con el código CASMO-4 para obtener una verificación de la metodología completa. Esta se ha probado usando un reactor de agua en ebullición del tipo BWR. Sin embargo, no hay ninguna preocupación o limitación respecto a su uso con otro tipo de reactor nuclear. Para la cuantificación de la incertidumbre se usa la metodología estocástica Gesellschaft für Anlagen und Reaktorsicherheit (GRS). Esta metodología hace uso del modelo de alta fidelidad y un muestreo no paramétrico para propagar la incertidumbre. Como resultado, el número de muestras (determinado con la fórmula revisada de Wilks) no depende del número de parámetros de entrada, sólo depende del nivel de confianza e incertidumbre deseados de los parámetros de salida. Además, las funciones de distribución de probabilidad no están limitadas a normalidad. El principal inconveniente es que se ha de disponer de las distribuciones de probabilidad de cada parámetro de entrada. Si es posible, las distribuciones de probabilidad de entrada se definen usando información encontrada en la literatura relacionada. En caso contrario, la incertidumbre se define en base a la opinión de un experto. Se usa un segundo escenario para propagar la incertidumbre de diferentes parámetros termo-hidráulicos a través del código acoplado TRACE5.0p3/PARCSv3.0. En este caso, se utiliza un reactor tipo PWR para simular un transitorio de una caída de barra. Como nueva característica, el núcleo se modela elemento a elemento siguiendo una discretización totalmente en 3D. No se ha encontrado ningún otro estudio que use un núcleo tan detallado en 3D. También se usa la metodología GRS y el DAKOTA 6.3 para este análisis de incertidumbre y sensibilidad.
Aquest treball de doctorat, desenvolupat a la Universitat Politècnica de València (UPV), té com a objectiu cobrir la primera fase del benchmark presentat pel grup d'experts Uncertainty Analysis in Modeling (UAM-LWR). La principal contribució al benchmark, per part de l'autor d'aquesta tesi, es el desenvolupament d'un programa de MATLAB sol¿licitat pels organitzadors del benchmark, el qual s'utilitza per a generar llibreries neutròniques a distribuir entre els participants del benchmark. El benchmark del UAM pretén determinar la incertesa introduïda pels codis multifísics i multiescala acoblats d'anàlisi de reactors d'aigua lleugera. El citat benchmark es divideix en tres fases: 1. Fase neutrònica: obtenir els paràmetres neutrònics i seccions eficaces del problema específic, col¿lapsats i homogeneïtzats, a més de la anàlisi de criticitat. 2. Fase de nucli: anàlisi termo-hidràulica i neutrònica per separat. 3. Fase de sistema: anàlisi termo-hidràulica i neutrònica acoblats. En aquesta tesi es completen els principals objectius de la primera fase. Concretament, es desenvolupa una metodologia per propagar la incertesa de les seccions eficaces i altres paràmetres neutrònics a través d'un codi lattice i un simulador de nucli. Es porta a terme una anàlisi d'incertesa i sensibilitat per a les seccions eficaces contingudes en la llibreria neutrònica ENDF/B-VII. La seua incertesa es propaga a través del codi lattice SCALE6.2.1, incloent les fases per col¿lapsar i homogeneïtzar, fins aplegar a la generació d'una llibreria neutrònica específica del problema. Després, la incertesa continguda en la esmentada llibreria pot continuar propagant-se a través d'un simulador de nucli, per a aquest estudi PARCSv3.2. Per a l'anàlisi d'incertesa i sensibilitat s'ha utilitzat el mòdul SAMPLER -disponible a l'última versió de SCALE- i la ferramenta estadística DAKOTA 6.3. Com a part d'aquest procés, també es desenvolupa una metodologia per a obtenir llibreries neutròniques en format NEMTAB per ser utilitzades en simuladors de nucli. S'ha realitzat una comparació amb el codi CASMO-4 per obtenir una verificació de la metodologia completa. Aquesta s'ha provat utilitzant un reactor d'aigua en ebullició del tipus BWR. Tanmateix, no hi ha cap preocupació o limitació respecte del seu ús amb un altre tipus de reactor nuclear. Per a la quantificació de la incertesa s'utilitza la metodologia estocàstica Gesellschaft für Anlagen und Reaktorsicherheit (GRS). Aquesta metodologia fa ús del model d'alta fidelitat i un mostreig no paramètric per propagar la incertesa. Com a resultat, el nombre de mostres (determinat amb la fórmula revisada de Wilks) no depèn del nombre de paràmetres d'entrada, sols depèn del nivell de confiança i incertesa desitjats dels paràmetres d'eixida. A més, las funcions de distribució de probabilitat no estan limitades a la normalitat. El principal inconvenient és que s'ha de disposar de les distribucions de probabilitat de cada paràmetre d'entrada. Si és possible, les distribucions de probabilitat d'entrada es defineixen utilitzant informació trobada a la literatura relacionada. En cas contrari, la incertesa es defineix en base a l'opinió d'un expert. S'utilitza un segon escenari per propagar la incertesa de diferents paràmetres termo-hidràulics a través del codi acoblat TRACE5.0p3/PARCSv3.0. En aquest cas, s'utilitza un reactor tipus PWR per simular un transitori d'una caiguda de barra. Com a nova característica, cal assenyalar que el nucli es modela element a element seguint una discretizació totalment 3D. No s'ha trobat cap altre estudi que utilitze un nucli tan detallat en 3D. També s'utilitza la metodologia GRS i el DAKOTA 6.3 per a aquesta anàlisi d'incertesa i sensibilitat.¿
Mesado Melia, C. (2017). Uncertainty Quantification and Sensitivity Analysis for Cross Sections and Thermohydraulic Parameters in Lattice and Core Physics Codes. Methodology for Cross Section Library Generation and Application to PWR and BWR [Tesis doctoral no publicada]. Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/86167
TESIS
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Twumasi, Afriyie Ebenezer. "Preparation and Evaluation of New Nanoporous Silica Materials for Molecular Filtration and for Core Materials in Vacuum Insulation Panels." Doctoral thesis, KTH, Byggnadsteknik, 2013. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-120330.

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Nanoporous materials for gas purification and thermal insulation have been studied and developed for application in many areas. It is known that a single adsorbent may not adequately control multiple contaminants. Further the utilization of nanoporous material as thermal insulator in building applications is limited due to high cost. Moreover, in view of the global environmental movement for clean air and reduction of heating energy consumption in built environment, the development of new and better nanoporous materials will not only facilitate major advances in gas adsorption and thermal insulation technology, but also meet the new challenges that cannot be met with the nanoporous materials that are currently available. This thesis presents a synthesis of new nanoporous silica based materials, and the characterization and application of these materials for molecular filtration and thermal insulation. Commercial nanoporous materials have been used for benchmarking for the pore properties, the applicability, and the performance of these new materials. First a double metal-silica adsorbent has been synthesized. The preparation procedure is based on the use of sodium silicate coagulated with various ratios of magnesium and calcium salts which yields micro-meso porous structures in the resulting material. The results show that molar ratios of Mg/Ca influence the pore parameters as well as the structure and morphology. The bimodal pore size can be tailored by controlling the Mg/Ca ratio. In the second synthesis, pure mesoporous silica, SNP has been prepared using glycerol as pore forming agent and monovalent salts as coagulant. This leads to material with large surface area and uniformed pore size centred at 43 or 47 nm.  The materials further exhibits a low bulk density in the range of 0.077 to 0.122 g/ml and possess a high porosity in the range of 95-97%. The influence of acid type (organic or inorganic) on the pore parameters and on the tapped density has also been investigated.   A synthesis method has also been developed for the preparation of carbon-silica composites. The method involves a number of routes, which can be summarised as addition of activated carbon particles to (I) the paste, (II) the salt solution, or (III) with the sodium silicate solution. In route II and III the activated carbon is present before coagulation. The routes presented here leads to carbon-silica composites possessing high micro porosity, meso porosity as well as large surface areas. The results further shows that pore size distribution may be tailored based on the route of addition of the carbon particles. Following route I and III a wide pore size (1-30 nm) was obtained whereas by route II a narrow pore size (1-4 nm) was observed.     MgCa-silica chemisorbents were also developed using either potassium hydroxide or potassium permanganate as impregnate chemicals. A direct or post-impregnation procedure was employed. The results revealed that the impregnate route and amount cause a reduction in both specific surface area and pore volume. Finally the thermal conductivity and dynamic adsorption of H2S, SO2 andtoluene were measured. Results show that at room temperature and atmospheric pressure, a thermal conductivity of 28.4 and 29.6 mW/m.K were obtained for the SNP mesoporous silicas. The dynamic adsorption behaviour of the chemisorbents and composites indicate their ability to absorbed H2S, SO2 andtoluene respectively. The highest H2S uptake corresponds to chemisorbents with 11.2-13.6 wt% KMnO4. The effect of impregnation route, amount of KMnO4 and its location in the pore system are likely the key factors in achieving a large H2S uptake. For SO2 adsorption, the highest uptake capacity was observed for MgCa-68/32-KOH. The results further suggest that the key to large SO2 uptake is as a result of the synergetic effect between large mesopore diameter and extensive mesopore volumes. Carbon-silica composites with carbon content 45 wt % exhibits high toluene adsorption with composite via route I having the highest toluene adsorption capacity (27.6 wt % relative to carbon content). The large uptake capacity of this composite was attributed to the presence of high microporosity volume and a wide (1-30 nm) bimodal pore system consisting of extensive mesopore channels (2-30 nm) as well as large surface area. These capacity values of carbon-silica composites are competitive to results obtained for commercial coconut based carbon (31 wt %), and better than commercial alumina-carbon composite (9.5 wt %).

QC 20130408

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Di, Chicco Augusto. "Optimization of a calculation scheme through the parametric study of effective nuclear cross sections and application to the estimate of neutronic parameters of the ASTRID fast nuclear reactor." Master's thesis, Alma Mater Studiorum - Università di Bologna, 2018.

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This thesis presents the project for the optimization of the APOLLO3® neutronic calculation scheme applied to the 4th generation fast neutron reactor ASTRID. APOLLO3® is the new multipurpose neutronic platform developed by the CEA. It incorporates many of the previous generation codes used in the French reactor core design supply chain. Like all deterministic codes, APOLLO3® solves the neutron transport equation with a discretization of the variables of interest: multi-group method for the energy, discrete ordinates and spherical harmonics for the angular variable, collision probabilities and characteristics methods for the spatial variable. The resolution of the transport equation handles useful quantities such as the neutron flux and multiplication factor, fission rates and cross sections to understand the physical behaviour of the reactor core. Currently it is not possible to use deterministic codes to simulate an entire reactor with a heterogeneous 3D geometry and a fine energy description, so to simplify the study of complete neutron field at core level, the calculation scheme is divided into two phases: lattice and core calculation. The main purpose of this work is to find an optimal degree of approximations of the calculation scheme for the evaluation of a desired physical effect and of the user constraints. In order to reach this optimum, several studies have been carried out with different levels of approximations. The results have been benchmarked with the ones obtained using the stochastic code TRIPOLI4®, used as a reference and to ensure a good accuracy. Furthermore, several sensitivity studies have been carried out to understand how the different approximations affect the macroscopic cross sections evaluation, because these dependences are not yet fully understood.
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Books on the topic "Core parameters"

1

Razzano, Giovanna. Il parametro delle norme non scritte nella giurisprudenza costituzionale. Milano: Giuffrè, 2002.

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Dembowski, Mary Ann. An evaluation of parameters influencing jet mixing using the WIND Navier-Stokes Code. Cleveland, Ohio: National Aeronautics and Space Administration, Glenn Research Center, 2002.

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Licci, Giorgio. Ragionevolezza e significatività come parametri di determinatezza della norma penale. Milano: A. Giuffrè, 1989.

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Robbins, Frederick W. Lumped-parameter interior ballistic computer code using the TTCP (The Technical Cooperation Program) model. Aberdeen Proving Ground, Md: Ballistic Research Laboratory, 1988.

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T, Urbanik. The sensitivity of evacuation time estimates to changes in input parameters for the I-DYNEV computer code. Washington, DC: Division of Radiation Protection and Emergency Preparedness, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, 1988.

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Bizyuk, Aleksandr. Fundamentals of abnormal psychology. ru: INFRA-M Academic Publishing LLC., 2020. http://dx.doi.org/10.12737/974663.

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The textbook is a Supplement to the course of lectures given at the faculties of psychology, where one of the sections of this discipline is clinical psychology. This publication has been updated to reflect the 11th International classification of diseases, changes in which also affect the classification aspects of mental disorders. In order to implement the principle of consistency in mastering knowledge of pathopsychology, the material is given in the context of General and clinical psychology, which facilitates the holistic assimilation of the specifics of this science and understanding its place among other related Sciences. In accordance with the requirements of didactics, the structuring of the material is based on the principle "from simple to complex"; at the end of each paragraph, test questions are offered, finding answers to which in the text of the book forms the core knowledge of the reader. The Chapter devoted to disorders of specific mental functions, in addition to General theoretical data, provides brief descriptions of psychodiagnostic techniques designed to assess the qualitative and quantitative parameters of recorded changes. When writing the book, we used a rich domestic and foreign material published in numerous sources. Meets the requirements of Federal state educational standards of higher education of the latest generation. It is intended for students of psychological, pedagogical and medical universities, primarily clinical psychologists, as well as for a wide range of specialists working in the information field of problems of the ratio of normal and altered psyche.
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Danckaert, Lieven. Changing EPP parameters. Oxford University Press, 2017. http://dx.doi.org/10.1093/oso/9780198759522.003.0005.

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This chapter starts with a description of the core facts concerning the VPAux/AuxVP alternation in the history of Latin. In the case of modal verbs and infinitives, there is a clear decline of the head-final order VPAux, whereas Late Latin BE-periphrases surprisingly prefer this order. Against the backdrop of these observations, the discussion then turns to the analysis of Classical and Late Latin clause structure. It is proposed that during the transition from Classical to Late Latin, a major parametric change took place related to the way the clausal EPP-requirement is satisfied. In the earlier grammar (‘Grammar A’), the entire VP undergoes A-movement to the high T-domain, resulting in the characteristic VPAux word order. In the later grammar (‘Grammar B’) the EPP-requirement is met by means of verb movement, with the VP staying in situ. In this grammar VPAux-orders are derived through roll-up movement, which is incompatible with the VOAux-pattern.
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Sethna, James P. Statistical Mechanics: Entropy, Order Parameters, and Complexity. 2nd ed. Oxford University Press, 2021. http://dx.doi.org/10.1093/oso/9780198865247.001.0001.

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This text distills the core ideas of statistical mechanics to make room for new advances important to information theory, complexity, active matter, and dynamical systems. Chapters address random walks, equilibrium systems, entropy, free energies, quantum systems, calculation and computation, order parameters and topological defects, correlations and linear response theory, and abrupt and continuous phase transitions. Exercises explore the enormous range of phenomena where statistical mechanics provides essential insight — from card shuffling to how cells avoid errors when copying DNA, from the arrow of time to animal flocking behavior, from the onset of chaos to fingerprints. The text is aimed at graduates, undergraduates, and researchers in mathematics, computer science, engineering, biology, and the social sciences as well as to physicists, chemists, and astrophysicists. As such, it focuses on those issues common to all of these fields, background in quantum mechanics, thermodynamics, and advanced physics should not be needed, although scientific sophistication and interest will be important.
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L, Korotev Randy, and United States. National Aeronautics and Space Administration., eds. Lithological variation with depth and decoupling of maturity parameters in Apollo 16 regolith core 68001/2. [Washington, DC: National Aeronautics and Space Administration, 1997.

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Institute Of Electrical and Electronics Engineers. IEEE C62.69A-2017 IEEE Standard for the Surge Parameters of Isolating Transformers used in Networking Devices and Equipment - Amendment 1: Addition of Saturated Core Secondary Winding Parameters. IEEE, 2017.

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Book chapters on the topic "Core parameters"

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Pyeon, Cheol Ho. "Neutron Spectrum." In Accelerator-Driven System at Kyoto University Critical Assembly, 125–56. Singapore: Springer Singapore, 2021. http://dx.doi.org/10.1007/978-981-16-0344-0_5.

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AbstractThe subcritical multiplication factor is considered an important index for recognizing, in the core, the number of fission neutrons induced by an external neutron source. In this study, the influences of different external neutron sources on core characteristics are carefully monitored. Here, the high-energy neutrons generated by the neutron yield at the location of the target are attained by the injection of 100 MeV protons onto these targets. In actual ADS cores, liquid Pb–Bi has been selected as a material for the target that generates spallation neutrons and for the coolant in fast neutron spectrum cores. The neutron spectrum information is acquired by the foil activation method in the 235U-fueled and Pb–Bi-zoned fuel region of the core, modeling the Pb–Bi coolant core locally around the central region. The neutron spectrum is considered an important parameter for recognizing information on neutron energy at the target. Also, the neutron spectrum evaluated by reliable methodologies could contribute to the accurate prediction of reactor physics parameters in the core through numerical simulations of desired precision. In the present chapter, experimental analyses of high-energy neutrons over 20 MeV are conducted after adequate preparation of experimental settings.
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Dub, Martin, and Vojtěch Dynybyl. "Sensor for In-time Identification of Deep Core Drilling Parameters." In Current Methods of Construction Design, 243–49. Cham: Springer International Publishing, 2019. http://dx.doi.org/10.1007/978-3-030-33146-7_28.

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Toro, Miguel, and Jessika Rojas. "Influence of Synthesis Parameters on Morphology, Crystalline Structure and Colloidal Stability of Core and Core-Shell LaPO4Nanoparticles." In Processing, Properties, and Design of Advanced Ceramics and Composites: Ceramic Transactions, 57–69. Hoboken, NJ, USA: John Wiley & Sons, Inc., 2016. http://dx.doi.org/10.1002/9781119323303.ch5.

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Cauli, Claudia, Magdalena Ortiz, and Nir Piterman. "Actions over Core-Closed Knowledge Bases." In Automated Reasoning, 281–99. Cham: Springer International Publishing, 2022. http://dx.doi.org/10.1007/978-3-031-10769-6_17.

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AbstractWe present new results on the application of semantic- and knowledge-based reasoning techniques to the analysis of cloud deployments. In particular, to the security of Infrastructure as Code configuration files, encoded as description logic knowledge bases. We introduce an action language to model mutating actions; that is, actions that change the structural configuration of a given deployment by adding, modifying, or deleting resources. We mainly focus on two problems: the problem of determining whether the execution of an action, no matter the parameters passed to it, will not cause the violation of some security requirement (static verification), and the problem of finding sequences of actions that would lead the deployment to a state where (un)desirable properties are (not) satisfied (plan existence and plan synthesis). For all these problems, we provide definitions, complexity results, and decision procedures.
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Hanafusa, H., Y. Takeuchi, and J. Noda. "Wavelength-Flattened Couplers Fabricated from Single-Mode Fibers with Different Core Parameters." In Springer Proceedings in Physics, 334–38. Berlin, Heidelberg: Springer Berlin Heidelberg, 1989. http://dx.doi.org/10.1007/978-3-642-75088-5_50.

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Zhu, Qisu, Shihua Zou, and Bing Yang. "Seismic Performance Analysis of RC Frame Core Tube Structure Considering Floor Parameters." In Lecture Notes in Electrical Engineering, 1567–73. Singapore: Springer Singapore, 2020. http://dx.doi.org/10.1007/978-981-15-5959-4_190.

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Yamanaka, Masao. "Sensitivity and Uncertainty of Criticality." In Accelerator-Driven System at Kyoto University Critical Assembly, 215–43. Singapore: Springer Singapore, 2021. http://dx.doi.org/10.1007/978-981-16-0344-0_8.

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AbstractExcess reactivity and control rod worth are generally considered important reactor physics parameters for experimentally examining the neutron characteristics of criticality in a core, and for maintaining safe operation of the reactor core in terms of neutron multiplication in the core. For excess reactivity and control rod worth at KUCA, as well as at the Fast Critical Assembly in the Japan Atomic Energy Agency, special attention is given to analyzing the uncertainty induced by nuclear data libraries based on experimental data of criticality in representative cores (EE1 and E3 cores). Also, the effect of decreasing uncertainty on the accuracy of criticality is discussed in this study. At KUCA, experimental results are accumulated by measurements of excess reactivity and control rod worth. To evaluate the accuracy of experiments for benchmarks, the uncertainty originated from modeling of the core configuration should be discussed in addition to uncertainty induced by nuclear data, since the uncertainty from modeling has a potential to cover the eigenvalue bias more than uncertainty by nuclear data. Here, to investigate the uncertainty of criticality depending on the neutron spectrum of cores, it is very useful to analyze the reactivity of a large number of measurements in typical hard (EE1) and soft (E3) spectrum cores at KUCA.
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Reshetova, Galina, Vladimir Cheverda, and Tatyana Khachkova. "Numerical Experiments with Digital Twins of Core Samples for Estimating Effective Elastic Parameters." In Communications in Computer and Information Science, 290–301. Cham: Springer International Publishing, 2019. http://dx.doi.org/10.1007/978-3-030-36592-9_24.

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Chen, Chang, Yi-jian Tao, Zhi-kun Zhu, and Zeng-qiang Cao. "Simulation Research on Core Control Parameters of Electromagnetic Force in Electromagnetic Riveting System." In Proceeding of the 24th International Conference on Industrial Engineering and Engineering Management 2018, 366–77. Singapore: Springer Singapore, 2019. http://dx.doi.org/10.1007/978-981-13-3402-3_39.

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Lyu, Yang. "Lead Cooled Fast Reactor Core Physics Parameters Calculation Based on Monte Carlo Method." In Proceedings of The 20th Pacific Basin Nuclear Conference, 783–91. Singapore: Springer Singapore, 2017. http://dx.doi.org/10.1007/978-981-10-2317-0_74.

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Conference papers on the topic "Core parameters"

1

Chen, Shu, Peng Ding, Shuowen Hu, Wenqing Xia, Min Liu, Fengwan Yu, and Wenhuai Li. "Optimization of Core Parameters Based on Artificial Neural Network Surrogate Model." In 2022 29th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2022. http://dx.doi.org/10.1115/icone29-90511.

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Abstract The methodology of artificial intelligence (AI), particularly artificial neural network (ANN), would be in favor of nuclear energy system development. These ANN simulators may provide more efficient means than the traditional nuclear design codes, especially for the design of the key parameters, such as core geometry and layout, material composition. In this paper, a neutronics calculation code SARAX and the corresponding multilayer perceptron (MLP) surrogate model were used as simulators for core parameters optimization of a reference lead based fast reactor. The pellet radius, enrichments and active height are the interested core parameters, and core burnup and power distribution are the target characteristics in the study. The training of structure and weight parameters in MLP network are based on about 5000 calculations of SARAX code. Test results of neural network show a good agreement between MLP surrogate model and SARAX code. The feasibility of the MLP surrogate model to be used in core parameters optimization was also discussed. Results showed that, the core surrogate model based on MLP could be quickly constructed and regulated, and be a more efficient simulators in a innovate reactor optimization. The above work is completed in Sinan Platform, a multidisciplinary intelligent design platform developed by China Nuclear Power Technology Research Institute Co. Ltd.
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Zhao, Nan, Lei Liao, Jiaming Li, Jinggang Peng, and Jinyan Li. "Analysis of parameters for high loss side core in chirally coupled core fiber." In International Conference on Optical Instruments and Technology 2015, edited by Jianqiang Zhu and Chunqing Gao. SPIE, 2015. http://dx.doi.org/10.1117/12.2192927.

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Santana-Abril, J., Y. Elejalde, J. M. Monzon-Verona, J. Sosa, and Juan A. Montiel-Nelson. "Effective Parameters for Ferrite-Core RID Tag Antennas." In 2018 IEEE 61st International Midwest Symposium on Circuits and Systems (MWSCAS). IEEE, 2018. http://dx.doi.org/10.1109/mwscas.2018.8623854.

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Teyssier, Christian, Donna L. Whitney, Megan Korchinski, and Patrice F. Rey. "METAMORPHIC CORE COMPLEXES: CRUSTAL-SCALE PARAMETERS CONTROLLING FOOTWALL BEHAVIORS." In GSA 2020 Connects Online. Geological Society of America, 2020. http://dx.doi.org/10.1130/abs/2020am-359159.

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Stoyka, Kateryna, Giulia Di Capua, Antonio Della Cioppa, Nicola Femia, and Giovanni Spagnuolo. "Identification of ferrite core inductors parameters by evolutionary algorithms." In 2015 IEEE 13th International Conference on Industrial Informatics (INDIN). IEEE, 2015. http://dx.doi.org/10.1109/indin.2015.7281981.

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Granados, Eduardo, Li-Jin Chen, Chien-Jen Lai, Kyung-Han Hong, and Franz X. Kärtner. "Wavelength scaling of hollow-core fiber compressor design parameters." In CLEO: Applications and Technology. Washington, D.C.: OSA, 2012. http://dx.doi.org/10.1364/cleo_at.2012.jth2a.18.

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Guo, Liancheng, and Andrei Rineiski. "Numerical Investigation of Corium Coolability in Core Catcher: Sensitivity to Modeling Parameters." In 2018 26th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/icone26-81841.

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To avoid settling of molten materials directly on the vessel wall in severe accident sequences, the implementation of a ‘core catcher’ device in the lower plenum of sodium fast reactor designs is considered. The device is to collect, retain and cool the debris, created when the corium falls down and accumulates in the core catcher, while interacting with surrounding coolant. This Fuel-Coolant Interaction (FCI) leads to a potentially energetic heat and mass transfer process which may threaten the vessel integrity. For simulations of severe accidents, including FCI, the SIMMER code family is employed at KIT. SIMMER-III and SIMMER-IV are advanced tools for the core disruptive accidents (CDA) analysis of liquid-metal fast reactors (LMFRs) and other GEN-IV systems. They are 2D/3D multi-velocity-field, multiphase, multicomponent, Eulerian, fluid dynamics codes coupled with a fuel-pin model and a space- and energy-dependent neutron kinetics model. However, the experience of SIMMER application to simulation of corium relocation and related FCI is limited. It should be mentioned that the SIMMER code was not firstly developed for the FCI simulation. However, the related models show its basic capability in such complicate multiphase phenomena. The objective of the study was to preliminarily apply this code in a large-scale simulation. An in-vessel model based on European Sodium Fast Reactor (ESFR) was established and calculated by the SIMMER code. In addition, a sensitivity analysis on some modeling parameters is also conducted to examine their impacts. The characteristics of the debris in the core catcher region, such as debris mass and composition are compared. Besides that, the pressure history in this region, the mass of generated sodium vapor and average temperature of liquid sodium, which can be considered as FCI quantitative parameters, are also discussed. It is expected that the present study can provide some numerical experience of the SIMMER code in plant-scale corium relocation and related FCI simulation.
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Koszelew, Jolanta, Slawomir Koszelew, and Andrzej Rusaczyk. "Analysis of the influence of core parameters on light propagation in ring-core optical fiber." In SPIE Proceedings, edited by Ryszard S. Romaniuk. SPIE, 2004. http://dx.doi.org/10.1117/12.568925.

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Yang, Wen, Fei Chao, Jinrong Qiu, Xing Li, and Baolin Liu. "Verification of PWR-Core Analysis Code CORAL Using VERA Core Physics Benchmark." In 2021 28th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2021. http://dx.doi.org/10.1115/icone28-64721.

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Abstract Three dimensional PWR-core analysis code CORAL developed by Wuhan Second Ship Design and Research Institute, which provides all functions required by PWR-core analysis calculation. These functions are neutron diffusion within the core and reflector, macroscopic depletion or microscopic depletion calculation analysis, multi-channel or sub-channel thermal-hydraulic analysis, one-dimensional heat transfer from nuclear fuel to the coolant, critical search by boron concentration or control rod position, integral and differential worth of neutron absorbers, neutron kinetics parameters for transient analysis, in-core neutron detector response simulation etc. CORAL is convenient to update and maintain in consider of modular, object-oriented programming technology. In order to verify the computational capabilities of the reactor core analysis software and methods, the US CASL project proposed the VERA reactor physics benchmark problem. This benchmark problem is based on the initial loading of Watts Bar Nuclear unit 1 as a model ranging from a simple 2D pin problem cell to the full cycle depletion and refueling of problem a 3D reactor core configuration. Aiming at the VERA benchmark problems, physics calculations are performed using the CORAL code, and the results of effective multiplication factor, assembly power distribution, control rod worth value and reactivity coefficient are obtained. By comparing with the KENO results provided in the benchmark problems, the calculation results of CORAL code are in good agreement. This shows that the CORAL code has the ability to calculate from 2D lattice to the 3D reactor core, and its computational accuracy is basically equivalent to the KENO code.
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Li, Chun-Yen, Rong-Jiun Sheu, Jinn-Jer Peir, Der-Sheng Chao, and Jenq-Horng Liang. "Effects of Various Fuel Parameters on the HTTR Criticality Calculations." In 2012 20th International Conference on Nuclear Engineering and the ASME 2012 Power Conference. American Society of Mechanical Engineers, 2012. http://dx.doi.org/10.1115/icone20-power2012-54857.

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The high temperature engineering test reactor (HTTR) in Japan is a 30 MWth helium-cooled and graphite-moderated reactor that was built to demonstrate the technological feasibility and advantages for the next-generation high-temperature gas-cooled reactors. The multi-level heterogeneous configuration of the HTTR core represents challenges to reactor physics calculations. A suitable core model is of vital significance for reliable neutronic analyses of the HTTR. In this study, a detailed HTTR core model is constructed based on the MCNP5 Monte Carlo code and the ENDF/B-VII neutron cross-section library. For code and model validation, several HTTR benchmark calculations were performed first and compared with the experimental results. After that, a variety of fuel design was explored to investigate their effects on the HTTR criticality calculations. The fuel design or parameters under consideration include the moderator-to-fuel ratio in the core, fuel rod dimension, and different fissile materials. The resultant trend of change in system multiplication factor and neutron spectrum was presented and discussed in this paper.
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Reports on the topic "Core parameters"

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Hanson, A. L., and D. Diamond. Calculation of Design Parameters for an Equilibrium LEU Core in the NBSR. Office of Scientific and Technical Information (OSTI), September 2011. http://dx.doi.org/10.2172/1030634.

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Hu, K., and D. R. Issler. A comparison of core petrophysical data with well log parameters, Beaufort-Mackenzie Basin. Natural Resources Canada/ESS/Scientific and Technical Publishing Services, 2009. http://dx.doi.org/10.4095/247449.

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Denman, Matthew R., and Dusty Marie Brooks. Fukushima Daiichi Unit 1 Uncertainty Analysis-Exploration of Core Melt Progression Uncertain Parameters-Volume II. Office of Scientific and Technical Information (OSTI), August 2015. http://dx.doi.org/10.2172/1211553.

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Torrijos, Ivan Dario Pinerez, Tina Puntervold, Skule Strand, Panagiotis Aslanidis, Ingebret Fjelde, and Aleksandr Mamonov. Core restoration: A guide for improved wettability assessments. University of Stavanger, November 2021. http://dx.doi.org/10.31265/usps.198.

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The initial wetting of a reservoir sets a limit for the EOR potential during water-based recovery operations and “Smart Water” injection. For this reason, an improved understanding of the factors influencing the wetting can help to control and better forecast oil production during water-based floods. To preserve and reproduce the original reservoir wettability is a challenging task and wrong cleaning and core restoration procedures can lead to incorrect wettability estimations and thus induce serious errors when evaluating the initial wettability of a reservoir system or its EOR potential by water-based methods. Thereby, there is a need to improve the chemical knowledge on interactions among the rock, brine and fluids present in reservoir systems. This will help to understanding the influence of the parameters affecting wettability during cleaning and core restoration processes. Understanding which are the main parameters influencing oil recovery processes is of great relevance. The objective of this document is to provide suggestions for added-value experiments, complementing and challenging the standard RCA and SCAL procedures, prior to performing experimental research in which wettability and wettability alteration processes are important. Lessons learned will be highlighted and new ideas to optimize core restoration protocols to preserve and closely reproduce wettability are put forward. These recommended practices target core restoration procedures after the core material has been received in the laboratory. The target audience for this document is engineers and scientists with an interest in core preparation for wettability studies.
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Jettestuen, Espen, Olav Aursjø, Jan Ludvig Vinningland, Aksel Hiorth, and Arild Lohne. Smart Water flooding: Part 2: Important input parameters for modeling and upscaling workflow. University of Stavanger, November 2021. http://dx.doi.org/10.31265/usps.200.

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This document presents some guidelines on how to conduct numerical investigations of the physicochemical effects of Smart Water flooding on different length scales. The National IOR Centre of Norway (NIORC) has developed several simulation tools. The objective of this report is to describe how three NIORC-developed simulation tools BADChIMP, IORCoreSim, and IORSim, can be used to investigate Smart Water effects on different length scales. We present which input parameters are needed by the simulation tools, and we discuss which processes these tools are suited to study. When working with different length scales, one of the challenges is how to upscale results obtained from smaller scales, i.e., pore and core scale experiments or simulations, to the field scale. Here, three relevant questions are: 1) how far do the Smart Water effects propagate into a reservoir? 2) What is the effect of reservoir temperature on Smart Water behavior? 3) How is the oil release, observed on core scale, related to the oil production from a field? This document targets research scientists planning to perform either pore scale simulations, core scale simulations, or field scale simulations for Smart Water studies. The technical level of the document is targeting an industry engineer.
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Hanson A. L. and Diamond D. Calculation of Design Parameters for an Equilibrium LEU Core in the NBSR using a U7Mo Dispersion Fuel. Office of Scientific and Technical Information (OSTI), June 2014. http://dx.doi.org/10.2172/1134663.

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Lee, C. H., H. K. Joo, W. S. Yang, and T. A. Taiwo. Implementation of nodal equivalence parameters in DIF3D-VARIANT for core analysis of prismatic Very High Temperature Reactor (VHTR). Office of Scientific and Technical Information (OSTI), March 2007. http://dx.doi.org/10.2172/925340.

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Hu, K. Petrophysical data from core samples in the "Mackenzie Corridor", Northwest and Yukon Territories: estimates of petroleum reservoir parameters. Natural Resources Canada/ESS/Scientific and Technical Publishing Services, 2009. http://dx.doi.org/10.4095/248214.

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Pavlovitchev, A. M. Kinetics Parameters of VVER-1000 Core with 3 MOX Lead Test Assemblies To Be Used for Accident Analysis Codes. Office of Scientific and Technical Information (OSTI), March 2000. http://dx.doi.org/10.2172/814131.

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Meeker, Jessica. Mutual Learning for Policy Impact: Insights from CORE. Sharing Experience and Learning on Approaches to Influence Policy and Practice. Institute of Development Studies (IDS), August 2021. http://dx.doi.org/10.19088/core.2021.005.

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On 23 June 2021, Southern Voice and the Institute of Development Studies co-hosted an online dialogue which aimed to enhance efforts to inform and influence policy by sharing learning between CORE projects, at different stages in their policy engagement activities, on their approaches and experiences at sub-national, national, and regional levels. The event was attended by over 70 participants from across the CORE cohort and highlighted the experiences of CORE partners, Partnership for Economic Policy (PEP), International Centre for Research on Women (ICRW), and Group for the Analysis of Development (GRADE). This learning guide captures the practical insights and advice from the event to help inform the practice of both participants and other projects across the portfolio. The guide is structured around the key challenges identified in influencing policy, particularly within the changing parameters of the current pandemic, highlighting key messages and examples from the three partners.
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