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    Wyświetlanie 1-3 z 3
    Tytuł:
    An application of LSC method for the measurement of gross alpha and beta activities in spiked water and drinking water samples
    Autorzy:
    Çakal, G. Ö.
    Güven, R.
    Yücel, H.
    Powiązania:
    https://bibliotekanauki.pl/articles/146427.pdf
    Data publikacji:
    2015
    Wydawca:
    Instytut Chemii i Techniki Jądrowej
    Tematy:
    alpha activity
    beta activity
    drinking water
    liquid scintillation counting (LSC)
    pulse shape analysis (PSA) calibration
    quench
    Opis:
    In this study, after the pulse shape calibration of a liquid scintillation counting (LSC) spectrometer (Quantulus 1220), the efficiency was determined depending on sample quenching parameters. Then, gross alpha and beta activities in two spiked water samples obtained from International Atomic Energy Agency (IAEA) were used for the validation of the ASTM D7283-06 method, which is a standard test method for alpha and beta activity in water by LSC. Later, the drinking water samples (35 tap water and 9 bottled water) obtained from different districts of Ankara, Turkey, were measured. The maximum gross alpha activities are measured to be 0.08 Bq/L for tap waters and 0.13 Bq/L for bottled waters, whereas the maximum gross beta activities are found to be 0.18 Bq/L for tap waters and 0.16 Bq/L for bottled waters. These results indicate that these drinking water samples are below the required limits, which are 0.1 Bq/L for alpha emitting radionuclides and 1 Bq/L for beta emitting radionuclides. As a result, gross alpha and beta activities in drinking water of Ankara were determined accurately by this validated LSC method. It is also worth noting that LSC is a rapid and accurate method for the determination of gross alpha and beta activities without requiring a tedious sample preparation.
    Źródło:
    Nukleonika; 2015, 60, No. 3, part 2; 637-642
    0029-5922
    1508-5791
    Pojawia się w:
    Nukleonika
    Dostawca treści:
    Biblioteka Nauki
    Artykuł
    Tytuł:
    235U isotopic characterization of natural and enriched uranium materials by using multigroup analysis (MGA) method at a defined geometry using different absorbers and collimators
    Autorzy:
    Yücel, H.
    Yeltepe, E.
    Yüksel, A. Ö.
    Dikmen, H.
    Powiązania:
    https://bibliotekanauki.pl/articles/146510.pdf
    Data publikacji:
    2015
    Wydawca:
    Instytut Chemii i Techniki Jądrowej
    Tematy:
    nuclear safeguards
    enrichment measurements
    uranium
    multigroup analysis
    enrichment meter principle
    Opis:
    Characterization of nuclear materials is an important topic within the context of nuclear safeguards, homeland security and nuclear forensics. This paper deals with the performance of multigroup gamma-ray analysis (MGA) method using the X- and γ-rays in the 80–130 keV region and enrichment meter principle (EMP) based on the analysis of 185.7 keV peak for a certain geometry using different absorbers and collimators. The results from MGA and those of EMP are compared. In particular, the effect of aluminum/lead absorbers and lead collimator on the enrichment determination of 235U in natural and low enriched samples is investigated in a given source- -detector geometry. The optimum diameter/height ratio for the Pb-collimator is found to be Dc/Hc = 1.4–1.6 in the chosen geometry. In order to simulate the container walls, ten different thicknesses of Al-absorbers of 141 to 840 mg·cm–2 and six different thicknesses of Pb-absorbers of 1120–7367 mg·cm–2 are interposed between sample and detector. The calibration coefficients (% enrichment/cps) are calculated for each geometry. The comparison of the MGA and EMP methods shows that the enrichment meter principle provides more accurate and precise results for 235U abundance than those of MGA method at the chosen geometrical conditions. The present results suggest that a two-step procedure should be used in analyses of uranium enrichment. Firstly MGA method can be applied in situ and then EMP method can be used at a defined geometry in laboratory.
    Źródło:
    Nukleonika; 2015, 60, No. 3, part 2; 615-620
    0029-5922
    1508-5791
    Pojawia się w:
    Nukleonika
    Dostawca treści:
    Biblioteka Nauki
    Artykuł
    Tytuł:
    Investigation of thermal neutron detection capability of a CdZnTe detector in a mixed gamma-neutron radiation field
    Autorzy:
    Yücel, H.
    Narttürk, R. B.
    Zümrüt, S.
    Gedik, G.
    Karadag, M.
    Powiązania:
    https://bibliotekanauki.pl/articles/146914.pdf
    Data publikacji:
    2018
    Wydawca:
    Instytut Chemii i Techniki Jądrowej
    Tematy:
    neutron detection
    CdZnTe
    prompt gamma ray
    thermal neutron
    cadmium
    neutron sensitivity
    241Am-Be source
    Opis:
    The aim of this study was to investigate the thermal neutron measurement capability of a CdZnTe detector irradiated in a mixed gamma-neutron radiation fi eld. A CdZnTe detector was irradiated in one of the irradiation tubes of a 241Am-Be source unit to determine the sensitivity factors of the detector in terms of peak count rate (counts per second [cps]) per neutron flux (in square centimeters per second) [cps/neutron•cm–2•s–1]. The CdZnTe detector was covered in a 1-mm-thick cadmium (Cd) cylindrical box to completely absorb incoming thermal neutrons via 113Cd(n,γ) capture reactions. To achieve, this Cd-covered CdZnTe detector was placed in a well-thermalized neutron fi eld (f-ratio = 50.9 ± 1.3) in the irradiation tube of the 241Am-Be neutron source. The gamma-ray spectra were acquired, and the most intense gamma-ray peak at 558 keV (0.74 γ/n) was evaluated to estimate the thermal neutron fl ux. The epithermal component was also estimated from the bare CdZnTe detector irradiation because the epithermal neutron cutoff energy is about 0.55 eV at the 1-mm-thick Cd filter. A high-density polyethylene moderating cylinder box can also be fi tted into the Cd fi lter box to enhance thermal sensitivity because of moderation of the epithermal neutron component. Neutron detection sensitivity was determined from the measured count rates from the 558 keV photopeak, using the measured neutron fluxes at different irradiation positions. The results indicate that the CdZnTe detector can serve as a neutron detector in mixed gamma-neutron radiation fields, such as reactors, neutron generators, linear accelerators, and isotopic neutron sources. New thermal neutron filters, such as Gd and Tb foils, can be tested instead of the Cd filter due to its serious gamma-shielding effect.
    Źródło:
    Nukleonika; 2018, 63, 3; 59-64
    0029-5922
    1508-5791
    Pojawia się w:
    Nukleonika
    Dostawca treści:
    Biblioteka Nauki
    Artykuł
      Wyświetlanie 1-3 z 3

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