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1
Content available remote

A Simple Approach to Data Analysis for the Detection of Hazardous Materials by Means of Neutron Activation Analysis

100%
Kaźmierczak Ł., Borsuk S., Gierlik M., Guzik Z., Iwanowska-Hanke J., Korolczuk S., Kozłowski T., Krakowski T., Marcinkowski R., Swiderski L., Szeptycka M., Szewiński J., Urban A.
Acta Physica Polonica A
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2015
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vol. 127
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issue 5
1540-1542
EN
Neutron activation analysis is a well-known method for determining isotopic composition of different materials. Due to the non-destructive nature of the method it can also be used in homeland security application, for example an inspection of packages of unknown origin. For this purpose hazardous material detection system (SWAN) was built. The device has an automatic decision algorithm that analyze the spectrum of gamma rays of ^{12}C, ^{14}N and ^{16}O. Characteristic ratio of these lines gives the requested answer. Using such a simple approach SWAN is able to obtain promising results.
2
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Determination of Natural Radioactivity and Associated Radiological Hazard in Excavation Field in Turkey (Oluz Höyük)

63%
Çetin B., Öner F., Akkurt İ.
Acta Physica Polonica A
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2016
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vol. 130
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issue 1
475-478
EN
The level of natural radioactivity may be varied with the increase in dependence of the depth into the earth. In present study, soil samples from different depths were collected in Oluz Höyük excavation area. The radionuclide concentrations in soil samples were determined by gamma-ray spectrometer which contains 3" × 3" NaI(Tl) detector connected to multichannel analyser. The photopeaks at 1460, 1764, and 2615 keV due to ⁴⁰K, ²²⁶Ra and ²³²Th, respectively, have been used. The obtained activity concentrations of ⁴⁰K, ²²⁶Ra and ²³²Th ranged from 656.03 to 1791.85 Bq/kg, 62.39 to 180.93 Bq/kg and 48.31 to 125.43 Bq/kg, respectively. To assess the radiological hazard of the natural radionuclides content in the soil samples of these area, the radium equivalent activities (the minimum value was 181.99 Bq/kg and the maximum value was 497.97 Bq/kg), the absorbed dose rate (the minimum value was 86.83 nGy/h and the maximum value was 237.22 nGy/h), annual effective dose rate (the minimum value was 0.11 mSv/y and the maximum value was 0.29 mSv/y) and external hazard index (the minimum value was 0.49 and the maximum value was 1.35) were calculated using measured activities.
3
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HK-Means: A Heuristic Approach to Initialize and Estimate the Number of Clusters in Biological Data

51%
Reddy Edla D., Gondlekar V., Gauns V.
Acta Physica Polonica A
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2016
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vol. 130
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issue 1
78-82
EN
K-means algorithm is one of the simplest and fastest clustering algorithms existing since more than four decades. One of the limitations of this algorithm is estimating number of clusters in advance. This algorithm also suffers from random initialization problem. This paper proposes a heuristic which initializes the cluster centers and estimates the number of clusters as a discrete value. The method estimates the number of clusters and initializes many cluster centers successfully for the clusters that are dense and separated significantly. The method selects a new cluster center in each iteration. The point selected is the point which is most dissimilar from the previously chosen points. The proposed algorithm is experimented on various synthetic data and the results are encouraging.
4
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Neutron-Induced Medical Radioisotope Production in a Conceptual Accelerator-Driven System, Fueled with Uranium Carbide

44%
Arslan A., Bakir G., Selçuklu S., Genç G., Yapici H.
Acta Physica Polonica A
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2016
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vol. 130
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issue 1
68-71
EN
In this study, the medical radioisotope production performance of a conceptual accelerator-driven system is investigated. Lead-bismuth eutectic is used as target material. The fuel core of the considered accelerator-driven system is divided into ten subzones, loaded with uranium carbide and various isotopes (isotopes of copper, gold, cobalt, holmium, rhenium, scandium, and thulium) and cooled with light water. As is known, light water is an effective moderator of neutrons as well as a good coolant. The fuel and the isotopes are separately placed as cylindrical rods with a cladding of carbon composite. The volume fractions of fuel, isotope, cladding and coolant are selected as 25%, 35%, 10% and 30%, respectively. The copper rods are placed into the first five subzones due to the fact that copper isotopes have low capture cross-section. In the case of the each radioisotope production, one of the other considered isotopes that have higher capture cross-section are placed into the following five subzones for optimization of fission, fissile breeding and radioisotope production. The graphite zone is located around the fuel core to reflect the escaping neutrons. Boron carbide (B₄C) is used as shielding material. In order to produce more neutrons (about 25-30 neutrons per 1 GeV proton), the target is irradiated with a continuous beam of 1 GeV protons. All neutronic computations have been performed with the high-energy Monte Carlo N-Particle Transport Code using the LA150 data library. The neutronic results obtained from these calculations show that the examined accelerator-driven system has a high neutronic capability, in terms of production of thermal power, fissile fuels, and medical radioisotopes.
5
Content available remote

Neutronic Analysis of LBE-Uranium Spallation Target Accelerator Driven System Loaded with Uranium Dioxide in TRISO Particles

44%
Bakir G., Selçuklu S., Genç G., Yapici H.
Acta Physica Polonica A
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2016
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vol. 130
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issue 1
30-32
EN
This study presents the neutronic performances of fissile breeding and energy production of a gas cooled accelerator-driven system with LBE-uranium dioxide (UO₂) spallation target. The accelerator-driven system is designed and optimized by considering various target materials, in terms of neutronic. Two different materials, LBE + natural UO₂ and LBE + 15% enrichment UO₂ are selected as target materials. The target zone is divided into two parts, one within the other; the outer part is pure LBE target part, and the inner part is UO₂ target part cooled with the helium gas. Tristructural-isotropic (TRISO)-coated fuel particles, containing UO₂ fuel, are embedded in a carbon matrix pebble with the packing fraction of a 29%, and the pebbles are placed in the UO₂ target part and in the fuel core with the packing fraction of a 60%. The fuel core is cooled with helium that is a high-temperature coolant. The target is bombarded with the continuous beams of a 1 GeV protons to produce high-flux neutrons that enter the fuel core. The fuel core is surrounded with a graphite reflector zone serving as both effective moderation and reflection of these neutrons. Furthermore, the whole system is enclosed by boron carbide, B₄C (shielding zone), to prevent the neutrons leakage out of the accelerator-driven system. The high-energy Monte Carlo code MCNPX along with the LA150 library is used for neutronic calculations. The numerical results bring out that the investigated accelerator-driven system has a high neutronic performance, from the energy production and fissile breeding points of view. Namely, it can be obtained over the thermal power of a 350 MW and produced over the fissile breeding of a 300 g/day.
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