UNDERWATER INSPECTION OF THE STATE OF ALUMINUM CONTAINERS AND THE STATE OF REACTOR RA SPENT FUEL PACKED IN THEM
by Milena MATAUŠEK, Zoran VUKADIN, Radojko PAVLOVIĆ, Ilija PLEĆAŠ, Obrad ŠOTIĆ, Snežana PAVLOVIĆ, Zoja IDJAKOVIĆ

Abstract: Actions performed in order to identify and improve storage conditions of spent fuel from RA research reactor are summarized. Recently accomplished inspection of sealed aluminum barrels, containing aluminum cladded metallic uranium fuel, are described in details. Based on the results of this inspection, options for future safe disposal of RA reactor spent fuel are proposed.

Key words: storage of spent nuclear fuel, research nuclear reactor, radiological and nuclear safety, aluminium barrel, uranium fuel

Note: Full text published in Serbian

STANDARD NEUTRON FIELDS IN GRAPHITE PRISM WITH Ra-a-Be SOURCE
by Vladan LJUBENOV, Miodrag MILOŠEVIĆ

Abstract: Neutron slowing-down and space-energy distribution of neutron flux in graphite prism with Ra-a-Be source were investigated experimentally, numerically and theoretically. Recent measurement and numerical experiment based on Monte Carlo calculations were carried out for the purpose of determining detailed space and energy distribution of neutron flux in the prism. Measurement of two-group neutron flux distribution along the central vertical axis of graphite prism was performed by irradiation of golden foils with cadmium covers and without them. Gamma and beta activities of foils were measured absolutely by using coaxial semiconductor germanium gamma detector and dual gas proportional alpha/beta detector, respectively. Starting from the results of numerical Monte Carlo experiment, neutron ages that enable analytical representation of neutron flux in broad energy intervals were determined. For the six chosen broad energy intervals the very good agreement between numerical experiment and Fermi age theory results was achieved with three age groups, corresponded source fractions and neutron ages.

Key words: graphite prism, neutron slowing-down, neutron flux, irradiation of foils, Monte Carlo, Fermi age

REFLECTION OF FAST ELECTRONS IN DP0 APPROXIMATION OF TRANSPORT EQUATION
by Jovan VUKANIĆ, Rodoljub SIMOVIĆ

Abstract: Reflection of high energy electrons from solids is treated by the approximative analytic solution of linearized transport equation. For the scattering of electrons on target atoms determined by screened Coulomb interaction and energy loss defined by Bethe-Bloch formula, the mean number of large angle deflections of an electron before slowing down to rest has been introduced. The approach is applicable in wide range of electron energy - from several tens of keV to several MeV - and for materials where the mean number of large angle deflections is large.
The isotropic approximation of collision integral is accepted and the Boltzmann equation in Laplace transformed form over relative path length is analytically treated by the ordinary DPN method. In the lowest order of approximation, we derived the expressions for energy distributions of backscattered electrons as well as particle and energy reflection coefficients. Comparison of our results with data of the computational bipartition model is presented.

Key words: electron backscattering, multiple collisions, transport equation, DPN method, reflection coefficients

SEMIGROUPS AND TRANSLATION INVARIANCE OF RADIATION FIELDS
by Velimir STANČIĆ

Abstract: Starting from a particle transport equation, considered as an operator dependent first order partial differential equation with respect to space-time coordinates, an answer will be given in this paper to a proposed question: what circumstances are relevant for the radiation field being translation invariant. The analytical forms of semi groups generated by linear Boltzmann transport operator are used for this purpose. The obtained transformed transport equation will be treated as first order abstract Cauchy problem with respect to space-time coordinates. In this way equivalence between transport equation and ordinary partial differential equation is being established. It is shown that this equation, regardless different approaches of its solving, satisfies the translation invariance, in a Hilbert space with coarser topology. This space is obtained as a direct integral over Case eigen- function spectrum of Hilbert spaces tensor product with respect to angular and space coordinates.

Key words: transport equation, semi groups, analytic solution, translation invariance, Fourier transformation, Cauchy problem, radiation fields, Hilbert spaces

HEAT GENERATION AND CORRESPONDING RISE IN TEMPERATURE DUE TO ABSORPTION OF THERMAL NEUTRONS
by Magda A. IBRAHIM

Abstract: The evaluation of the heat generation and the corresponding rise in temperature due to absorption of thermal neutrons are carried out for some selected materials (water, ordinary concrete, heavy concrete, paraffin wax, graphite, lead and iron). The absorbed thermal flux is estimated by a simplified solution of the diffusion equation. The heat generation is considered to be the product of the macroscopic absorption cross-section of thermal neutrons by the corresponding thermal flux. Material thermal conductivity and the calculated heat generation then estimate the investigated rise in temperature.

Key words: thermal neutrons, heat generation, temperature rise, diffusion equation

CALCULATION OF NEUTRON PRODUCTION IN SELECTED MATERIALS BY BEAM OF CHARGED PARTICLES OF INTERMEDIATE ENERGIES
by Milan PEŠIĆ

Abstract: A comparison study on calculation of neutron production by beam of protons and deuterons in different target materials, in energy range from 10 MeV to 75 MeV, is shown in the paper. An idealised cylindrical target is bombarded, along the central axis, per pendicularly at target base, by an infinite thin particle beam. Simulation is carried out for the target material surrounded by void, i.e. the "return effect" from possible surrounding materials in a real system is not encountered. The study is carried out using Monte Carlo based computer codes for intermediate and high-energy nucleon transport: LCS, ver. 2.7 (LANL, USA) and SHIELD (INR, Russia). Total number of emitted neutrons in 4p solid angle per incident particle (Yield) and spectrum of emitted neutrons escaping the target surfaces are determined for different targets materials: ²⁰⁸Pb/Pb, ²³⁸U/U, ¹⁸⁴W/W, Be and ⁷Li. Maximum neutron yield, near 30%, is calculated for proton beam energy of 75 MeV bombarding ²³⁸U/U target, shaped as mentioned above. Generally, neutron yield for deuteron beam is less than neutron yield for proton beam of the same energy for targets made from high-Z nuclides. The opposite conclusion is derived for the target made from low-Z nuclides materials.

Key words: simulation, ADS target, proton, deuteron, neutron yield, neutron spectrum

ET-RR-2 REACTOR SHIELDING CALCULATIONS
by Mahmoud M. IMAM, Magda A. IBRAHIM

Abstract: Shielding calculations were carried out for Egypt's second research reactor ET-RR-2 using the three dimensional Monte Carlo code MULTIKENO after being modified by us to account for gamma attenuation. In this concern, MULTIKENO was developed from a Monte Carlo code for only neutron attenuation to a code for both neutron and gamma attenuation in shielding materials. Results for the total dose rate, gamma and total flux through the different reactor shielding materials in the radial direction as calculated by the new developed MULTIKENO Monte Carlo code are compared with those given in the safety analysis report (SAR) of the reactor, which are calculated using the one dimensional transport code ANISN. This comparison proved that all modifications introduced in the Monte Carlo code MULTIKENO were justified.

Key words: research reactor, shielding calculation, Monte Carlo, gamma attenuation, three-dimensional geometry

FAST-BREEDER-POWER REACTOR RECORDS IN THE INIS DATABASE
by Nada MARINKOVIĆ

Abstract: This report presents a statistical analysis of more than 19,700 records of publications concerned with research and technology in the field of fast breeder power fission reactors which are included in the INIS Bibliographic Database for the period from 1970. to 1999. The main objectives of this bibliometric study were: to make an inventory of the fast breeder power reactor related records in the INIS Database; to provide statistics and scientific indicators for the INIS users, namely science managers, researchers, engineers, operators, scientific editors and publishers, decision-makers in the field of fast breeder power reactors related subjects; to extract other useful information from the INIS Bibliographic Database about articles published in fast breeder reactors research and technology.
The quantitative data in this report are obtained for various properties of relevant INIS records such as year of publication, secondary subject categories, countries of publication, language, publication types, literary types, etc.

Key words: fission reactor, fast breeder power reactor, bibliographic database, bibliometric study, INIS