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A compact and modern equipment for implementing the fast neutron and γ-ray transmission technique (FNGT) has been developed in order to estimate the sulphur content of crude oil. FNGT is employed for non- destructive analysis of different kinds of samples. The compact system presented in this work represents an improvement of our previous experimental set-up [1, 2]. It makes use of a Cf source, an EJ-301 liquid scintillator detector (2″ × 2″) with excellent n/γ discrimination capabilities, and modern nuclear electronics based on fast digitizers. The fast neutron and gamma transmission technique was employed to study a system for on-line sulphur concentration measurement in Venezuelan heavy sour oil. The range of sulphur concentrations investigated is between 0.1 and 6.5 wt%. The equipment performances and limitations are compared with those predicted by a Monte Carlo model built in GEANT4 v10.01. The results show the possibility to implement a compact unit for on-line determination of sulphur concentration in crude oil. [ABSTRACT FROM AUTHOR]

During electron beam stopping on natW target in a linear accelerator, photonuclear (γ, xn) reactions occur; the tungsten converter provides a non-negligible neutron yield with an energy spectrum that significantly depends on surrounding mass nuclei. Reduction of the neutron radiation field is convenient to limit the side-effects that accompany the tumor or cancer radiotherapy. A close-in irradiation geometry is proposed to improve therapy effectiveness. The convenience of the proposed experimental arrangement is assessed using Monte Carlo simulation and experimental results based on nuclear track-etch methodology. Photoneutron yield for two energy groups (thermal and epithermal) are determined experimentally via boron (98%) converter and cadmium-filter employing a passive detector (poly allyl di-glicol carbonate polimer). Etched track diameter histograms are described by distribution functions to determine the ratio between thermal and higher energy neutrons. New insights are given into therapy beam quality and radiotherapy dose delivery based on bar histograms unfolding. • New irradiation with close-in geometry to provide higher-efficiency of radiotherapy. • Monte Carlo simulation to assess radiation field effectiveness and acceptability. • Distribution function of charged particles in on nuclear track-etch methodology. • Fast neutron yield at the W nat target converter. [ABSTRACT FROM AUTHOR]

A higher yield of rare isotope production methods, for example, isotope separation on-line (ISOL), is expected to be developed for the EURISOL facility. In this paper as a part of the ongoing project, high power-target assembly and passive detector inclusion are given. Theoretical calculations of several configurations were done using Monte Carlo code FLUKA aimed to produce 1015 fiss/s on LEU-Cx target. The proposed radioactive ion beam (RIB) production relies on a high-power (4MW) multibody target; a complete target design is given. Additionally we explore the possibility to employ PADC passive detector as a complementary system for RIB characterization, since these already demonstrated their importance in nuclear interactions phenomenology. In fact, information and recording rare and complex reaction product or short-lived isotope detection is obtained in an integral form through latent track formation. Some technical details on track formation and PADC detector etching conditions complete this study. [ABSTRACT FROM AUTHOR]

This paper presents three new materials composed of Ti X V 0,035X Cu X-1,035 (X = 2, 4 and 6%), is proposed as Linear Accelerator target. Its response to electron beam based on photoneutron production, is assessed by MC simulation and nuclear track-etch methodology. The outcome is compared to a tungsten target irradiated by energetic 16 MeV electron beam. Photoneutron yield, of two energy groups (thermal and epithermal) were determined via converter matter 10B (98%) and Cd-filter by PADC-track density comparison. The multi-metal Ti 2 V 0,07 Cu 97.93 target related to therapy beam quality, resulted advantageous in comparison to that provided by W-target, commonly used in the LINAC. • New target is assessed by thermal and fast photo-neutron yield of a medical LINAC. • The MCNP Monte Carlo code provide expected photoneutron dose per target alloy of Cu, Ti, V. • Low cost PADC-detector is convenient to assess target photoneutron production. • Partially amorphous target related therapy beam quality is better in comparison to tungsten. [ABSTRACT FROM AUTHOR]

Abstract: The light output response and the neutron and gamma-ray detection efficiency are determined for liquid scintillator EJ-309. The light output function is compared to those of previous studies. Experimental efficiency results are compared to predictions from GEANT4, MCNPX and PENELOPE Monte Carlo simulations. The differences associated with the use of different light output functions are discussed. [Copyright &y& Elsevier]

Paraffin, water and water-extended polyester (WEP) were used as main moderator to design a mobile shielding for a 666 GBq 241Am–Be source used in oil industry. The shielding performance was estimated using Monte Carlo methods where the γ-rays induced by the neutron interaction with the shielding materials were also included. The spectra of neutrons and γ-rays around the shielding were estimated, as well as the total neutron and g-ray total fluences per history. The neutron source strength was used to calculate ambient dose equivalent rates, aiming to satisfy the international recommendation (2 mSv/h at the shielding surface). Moderators modify the neutron spectrum of the source reducing the amount of fast neutrons (0.5–11 MeV) and producing epithermal and thermal neutrons. During neutron transport in the moderator neutron capture and inelastic scattering produce gamma-rays. Paraffin has the best shielding performance however it is not suitable because can be affected by extreme temperatures. WEP and water have similar shielding performance, however water do not satisfy the international recommendations because at one site the neutron and γ-ray doses rates are larger than 2 mSv/h. WEP is the best option because fulfill the international recommendations, it is stable and temperatures where paraffin fails. The WEP-based shielding weights 66 kg. • Using Monte Carlo methods, a mobile shielding for a241Am–Be source was calculated. • The geometric parameters and type of materials were established. • With the H*(10) calculations it is possible to establish the most adequate shields. • A geometry with dimensions for the construction of the container is presented. [ABSTRACT FROM AUTHOR]