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Anelli M.,National Institute of Nuclear Physics, Italy | Bertolucci S.,National Institute of Nuclear Physics, Italy | Bini C.,University of Rome La Sapienza | Bini C.,National Institute of Nuclear Physics, Italy | And 28 more authors.
Nuclear Instruments and Methods in Physics Research, Section A: Accelerators, Spectrometers, Detectors and Associated Equipment | Year: 2011

The neutron detection efficiency of a sampling calorimeter made of 1 mm diameter scintillating fibers embedded in a lead/bismuth structure has been measured at the neutron beam of The Svedberg Laboratory at Uppsala. A significant enhancement of the detection efficiency with respect to a bulk organic scintillator detector with the same thickness is observed. © 2010 Elsevier B.V. All rights reserved. Source


Bilodid I.,Institute of Safety Research
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2010

Codes for reactor core calculations use few-group cross sections (XS) which depend on local burnup, given in terms of the energy produced per fuel mass (MWd/kgHM). However, a certain burnup value can be reached under different spectral conditions depending on moderator density and other local parameters. Neglecting these spectral effects, i.e. applying the summary-burnup value only, can cause considerable errors in the calculated power density. This paper describes a way to take into account spectral-history effects. It is shown that the respective XS correction linearly depends on the actual Pu-239 concentration. The applicability of the method was proved not only for usual uranium oxide fuel, but also for mixed uranium/plutonium oxide (MOX) and fuel assemblies with burnable absorber. The code DYN3D was extended by new subroutines which calculate the actual distribution of Pu-239 in the core and apply a spectral-history correction for the XS. Copyright © 2010 by ASME. Source


Da Silva M.J.,Federal Technological University of Parana | Da Silva M.J.,Institute of Safety Research | Hampel U.,Institute of Safety Research | Rodriguez I.H.,University of Sao Paulo | Rodriguez O.M.H.,University of Sao Paulo
6th World Congress in Industrial Process Tomography | Year: 2010

In this article, we report on the application of a novel wire-mesh sensor based on permittivity (capacitance) measurements to investigate the flow of viscous oil and water in a horizontal pipe. Holdup values were calculated from the raw data delivered by the wire-mesh sensor using different mixture permittivity models. Furthermore, these data were validated against quick-closing valve measurements showing that Maxwell-Garnett mixing model for permittivity are better suited for the case of investigated liquid-liquid disperse flow. Images of time-averaged cross-sectional holdup distribution were used to visualize and disclose some details of the flow. © International Society for Industrial Process Tomography, 2010. All rights reserved. Source


Da Silva M.J.,Federal Technological University of Parana | Da Silva M.J.,Institute of Safety Research | Fischer F.,Institute of Safety Research | Thiele S.,Institute of Safety Research | Hampel U.,Institute of Safety Research
6th World Congress in Industrial Process Tomography | Year: 2010

In this paper, two fast flow imaging modalities, ultrafast electron beam x-ray tomography and wire- mesh sensor, are comparatively evaluated. The x-ray scanner ROFEX and a 24 × 24 wire-mesh sensor operated by a capacitive electronics were applied to measure the flow in an experimental two- phase flow loop. Part of the flow loop is a two metre tall round pipe of 50 mm inner diameter operated with water and air under controlled conditions. Different flow patterns and a broad range of void fraction values were generated. The ROFEX scanner was installed to visualize the flow just underneath the wire-mesh sensor. Flow images as well as void fraction profiles were compared. Results have shown good agreement in the outputs generated by both techniques. Only for low gas flow rate (low void fraction) the wire-mesh sensor overestimates the void fraction. © 2014 International Society for Industrial Process Tomography. Source


Bilodid I.,Institute of Safety Research | Mittag S.,Institute of Safety Research
Annals of Nuclear Energy | Year: 2010

Codes for reactor core calculations use few-group cross sections (XS) which depend on local burnup, given in terms of the energy produced per fuel mass (MWd/kgHM). However, a certain burnup value can be reached under different spectral conditions depending on moderator density and other local parameters. Neglecting these spectral effects, i.e. applying the summary-burnup value only, can cause considerable errors in the calculated power density. This paper describes a way to take into account spectral-history effects. It is shown that the respective XS correction linearly depends on the actual Pu-239 concentration. The applicability of the method was proved not only for usual uranium oxide fuel, but also for mixed uranium/plutonium oxide (MOX) and fuel assemblies with burnable absorber. The code DYN3D was extended by new subroutines which calculate the actual distribution of Pu-239 in the core and apply a spectral-history correction for the XS. © 2010 Elsevier Ltd. All rights reserved. Source

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