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This work is dedicated to figuring out robust epoxy/magnetite/boron carbide (EP/Mag/B4C) composite for radiation attenuation at multiple applications related to nuclear installations, as well as restoration mortar for cracking developed in concrete biological shields. The mechanical properties (flexural, compressive, and impact strengths) and the physical properties (water absorption, porosity, and dry bulk density), each, have been performed to label the composite integrity for practical application. In practice, attenuation properties have been performed by using a collimated beam emitted from spontaneous fission 252Cf (100 μg) neutron source and neutron gamma spectrometer with stilbene scintillator. The pulse shape discrimination technique which would come of the zero cross over method was used to measure the fast neutron and gamma-ray spectra. Thermal neutron fluxes have been measured by using the thermal neutron detection system and the BF-3 detector. The attenuation parameters: precisely, macroscopic effective removal cross-sections ΣR (cm-1), macroscopic cross-sections Σ (cm-1), and total attenuation coefficients μ (cm-1) of fast and thermal neutrons and total gamma-rays respectively were evaluated using the attenuation relations. Also, the MCNP5 code and MERCSF-N program have been used to compute the parameters theoretically. When applicable, measured and calculated results were compared, and it tells us a comprehensive agreement.

Advertising play a n important role in promoting goods and services, and impacting on purchasing decision, consuming behavior, social and moral behavior of the recipient. The effectiveness of advertising is sometimes affected by the extent of advertiser ethical commitment towards both of competitors and recipients, and commitment to ethical principles, standards which governs the form of competition between both of them and affects the mental image of products and the interaction of the recipient with it Since advertising’s first goal is Directing the competitive advertising message to the recipient in the appropriate timing to reach maximum degrees of persuasion. whereas Compliance with the advertising covenants and its laws is extremely important to ensure the effectiveness of these competitive ads ensuring the respect of recipients for both competing parties. The use of competitive advertising has developed recently especially ads that show competitor negatively and deliberately highlight the weakness in the competing products either directly or implicitly. Hence the study dealt with studying the ethics of advertising in competitive advertising and its nature and characteristics. Methodology: analytical descriptive methodology Statement of the problem: How can advertising ethics be used for ruling standards for competitive advertising? Objectives: The research aims to study the negatives and positives of competitive advertising, its characteristics,mechanisms of it’s application, and concluding ethical standards for competitive advertising. Results: The research concluded that: Competitive advertising is an advertising technique in which the advertiser claims preference and surpasses of the products which he is promoting to over the interestof his competitors by using direct and indirect compactivity. Also competitive advertising should reflect ethical principles and values, not to be misleading and do not cause controversy on generally agreed ethical values. commitment to ethical values in competition between institutions Helps to activate them positively and improve the mental image of the institution which recipient has. Competitive advertising helps in forcing brands to improve the quality of their products as well as been a strategy which helps to reduce barriers and break into new markets.

The present work deals with the shielding properties of lead glass composites to find out its integrity for practical shielding applications and radiological safety. Composites of different lead oxide ratios ( x = 0, 5, 10, 15 and 25 wt.%) have been prepared by the Nasser Glass and Crystal Company (Egypt). Attenuation measurements have been carried out using a collimated emitted beam from a fission 252 Cf (100 μg) neutron source, and the neutron–gamma spectrometer with stilbene scintillator. The pulse shape discriminating (P.S.D.) technique based on the zero cross-over method was used to discriminate between neutron and gamma-ray pulses. Thermal neutron fluxes were measured using the BF3 detector and thermal neutron detection system. The attenuation relations were used to evaluate fast neutron macroscopic effective removal cross-section Σ R-Meas (cm −1 ), gamma rays total attenuation coefficient μ (cm −1 ) and thermal neutron macroscopic cross-section Σ Meas (cm −1 ). Theoretical calculations have been achieved using MCNP5 code to calculate the same two parameters. Also, MERCSF-N program was used to calculate fast neutron macroscopic removal cross-section Σ R-MER (cm −1 ). Measured and MCNP5 calculated results have been compared and were found to be in reasonable agreement.

In the study, the usability of epoxy/ilmenite (EP/IIm) composite material as a radiation shielding in many applications and a restoration/injection mortar for cracks developing in biological concrete shields was investigated. Radiation attenuation properties of EP/IIm composite material was searched out using 252 Cf (100 μg) neutron source and gamma spectrometer with stilbene organic scintillator. Fast neutron fluence rates and gamma fluxes were measured and displayed as energy distributions and attenuation relations. Thermal neutron fluence rates were measured using BF3 detector and displayed as attenuation relations. The experimental attenuation parameters; macroscopic effective removal cross-section Σ R (cm −1 ), total attenuation coefficient μ (cm −1 ) and macroscopic cross-section Σ (cm −1 ) of fast neutrons, gamma rays and thermal neutron respectively have been determined. Theoretical calculations have been achieved using MCNP-4C2 code to evaluate the concerned parameters. Also, MERCSF – N program had been used to recalculate Σ R (cm −1 ). The obtained experimental and theoretical results were compared, where a reasonable agreement was recognized.

A trial has been made to create insulating Epoxy/Barite (EP/Brt) ( ρ =2.85 g cm −3 ) and Crosslinked Unsaturated Polyester/Barite (CUP/Brt) ( ρ =3.25 g cm −3 ) composites with radiation attenuation and shielding capabilities. Experimental work regarding mechanical and physical properties was performed to study the composites integrity for practical applications. The properties were found to be reasonable. Radiation attenuation properties have been carried out using emitted collimated beam from a fission 252 Cf (100 µg) neutron source, and the neutron–gamma spectrometer with stilbene scintillator. The pulse shape discriminating (P.S.D) technique based on the zero cross-over method was used to discriminate between neutron and gamma-ray pulses. Thermal neutron fluxes, measured using the BF 3 detector and thermal neutron detection system, were used to plot the attenuation relations. The fast neutron macroscopic effective removal cross-section Σ R , gamma ray total attenuation coefficient µ and thermal neutron macroscopic cross-section Σ have been evaluated. Theoretical calculations have been achieved using MCNP-4C2 code to calculate Σ R , µ and Σ . Also, MERCSF-N program was used to calculate macroscopic effective removal cross-section Σ R . Measured and calculated results have been compared and were found to be in reasonable agreement.

This work is concerned with studying polyester/ magnetite CUP/Mag ( ρ =2.75 g cm −3 ) and polyester/ ilmenite CUP/Ilm ( ρ =2.7 g cm −3 ) composites for shielding of medical facilities, laboratory hot cells and for various purposes. Mechanical and physical properties such as compressive, flexural and impact strengths, as well as, a.c. electrical conductivity, specific heat, water absorption and porosity have been performed to evaluate the composite capabilities for radiation shielding. A collimated beam from fission 252 Cf (100 µg) neutron source and neutron-gamma spectrometer with stilbene scintillator based on the zero cross over method and pulse shape discrimination (P.S.D.) technique have been used to measure neutron and gamma ray spectra. Fluxes of thermal neutrons have been measured using the BF 3 detector and thermal neutron detection system. The attenuation parameters, namely macroscopic effective removal cross-section Σ R , total attenuation coefficient µ and macroscopic cross-section Σ of fast neutrons, gamma rays and thermal neutrons respectively have been evaluated. Theoretical calculations using MCNP–4C2 code was used to calculate Σ R , μ and Σ . Also, MERCSF-N program was used to calculate macroscopic effective removal cross-section Σ R . Measured and calculated results were compared and reasonable agreement was found.

An approximate method for calculating the attenuation of fast neutrons can be achieved by using the “macroscopic effective removal cross-section” concept, usually denoted by Σ R (cm −1 ). The conventional method for the quantitative determination of Σ R (cm −1 ) for fast neutrons, which has been applied for several composites based on manual calculations, has a number of restrictions; a limited number of investigated samples, a lot of time, the possibility of random errors, etc. This work deals with the development, validation and application of the macroscopic effective removal cross-section of fast neutrons [MERCSF-N] computer program for calculating Σ R (cm −1 ) for fast neutrons transmitted through homogeneous mixtures, composites, concretes and compounds. The required physical data representing all periodic table elements have been compiled on the basis of the recommended published data and stored in a data base file. The validity of MERCSF-N has been confirmed by comparing its predictions with previously published results, both manually calculated and measured, where a good agreement was found. In addition, the code is used for calculating the macroscopic effective removal cross-sections of steel alloy, aluminum and magnetite concrete, which have previously been measured and reported. The calculated and measured results were compared and a good agreement was noticed.

Investigations have continued for economic utilization of lignocellulosic waste, which represents one of the environmental problems facing third world countries. In this work, we intended to use old newsprint paper (ONP) as a filler in radiation shielding rubber composite. In this respect, the treatment of ONP with magnetite was carried out. The effect of incorporating untreated and treated newsprint in the rubber matrix on the swelling, electrical and mechanical properties, as well as gamma and neutron-shielding power has been studied. The changes in physico-mechanical properties of such composites by γ-irradiation were also investigated. This study led to conclusion that composite samples containing 6% wax and 54% newspaper or 18% wax and 42% modified ONP has the higher behavior to γ-irradiation dose. The incorporation of modified waste paper into the rubber matrix increases the conductivity and enhances the mechanical properties of produced composites. At higher non modified and modified newspaper loadings a serious loss of rubber elasticity were noticed. Composite with good attenuation properties for both gamma rays and thermal neutrons was produced from incorporation of 54% modified newsprint with 6% paraffin wax.

Trial was made to create composites from agriculture fibers, which have good mechanical, physical and radiation attenuation properties for different applications. Fast neutron and gamma ray spectra, as well as, slow neutron fluxes behind samples of cement–fiber (CF) ( ρ =2.095 g cm −3 ) and cement–fiber–magnetite (CFM) ( ρ =2.858 g cm −3 )/composites have been investigated. Neutron and gamma spectra have been measured using a collimated reactor beam and neutron- gamma spectrometer with stilbene scintillator. A pulse shape discrimination technique based on zero cross over method was used to discriminate between neutron and gamma ray pulses. While, slow neutron fluxes have been measured using a collimated reactor beam and BF 3 counter. Results were used to achieve removal cross sections (Σ R,meas. ), total attenuation coefficients ( μ ) and macroscopic cross sections (Σ) of fast neutrons, gamma rays and slow neutrons respectively from the attenuation relations. Also, removal cross sections (Σ R, cal. ) and total mass attenuation coefficients ( μ / ρ ) of fast neutrons and gamma rays have been calculated using the composites elemental composition and XCOM code respectively. Measured and calculated results were compared and a reasonable agreement was found.

Study has been made of the radiation shielding provided by recycled agricultural fibre and industrial plastic wastes produced as composite materials. Fast neutron and gamma-ray spectra behind composites of fibre–plastic (r ¼ 1.373 g cm � 3 ) and fibre–plastic–lead (r ¼ 2.756 g cm � 3 ) have been measured using a collimated reactor beam and neutron–gamma spectrometer with a stilbene scintillator. The pulse shape discriminating technique based on the zerocross-over method was used to discriminate between neutron and gamma-ray pulses. Slow neutron fluxes have been measured using a collimated reactor beam and BF3 counter, leading to determination of the macroscopic cross-section (S). The removal cross-sections (SR) of fast neutrons have been determined from measured results and elemental composition of the composites. For gamma-rays, total linear attenuation coefficients (m) and total mass attenuation coefficients (m=r) have been determined from use of the XCOM code and measured results. Reasonable agreement was found between measured and calculated results. r 2003 Elsevier Science Ltd. All rights reserved.

This work deals with the study of ilmenite/epoxy composite as an injecting mortar for cracks developed in biological concrete shields, as well as, neutrons and gamma rays attenuation. Effects of the particle size on the mechanical strengths have been studied for epoxy resin filled with crushed ilmenite with different maximum particle sizes ranging from 32 to 500 μm. Thermal neutrons and gamma rays attenuation in ilmenite/epoxy composites with 75 and 80 wt.% of ilmenite concentration have been investigated. The total mass attenuation coefficients μ / ρ (cm 2 g −1 ) of gamma ray for five ilmenite/epoxy composites have been calculated using the XCOM program (version 3.1) at energies from 10 keV to 100 MeV. Also, the total mass attenuation coefficients ( μ / ρ ) have estimated based on the measured total linear attenuation coefficients (μ) and compared with the calculated results where, a reasonable agreement was found.

In the present work the authors have studied the attenuation properties of baryte concrete as a biological shield for nuclear power plants, particle accelerators, research reactors, laboratory hot cells and different radiation sources. Investigation has been performed by measuring the transmitted fast neutron and gamma ray spectra through cylindrical samples of baryte concrete (ρ=3.49 gcm-3). A reactor-collimated beam and neutron-gamma spectrometer with stilbene scintillator were used during measurements. A pulse shape discriminating technique based on zero cross over method was used to discriminate between neutron and gamma pulses. Removal cross-section ΣR,eff and total attenuation coefficient µeff of neutrons and gamma rays were evaluated from the obtained results. Also, slow neutron fluxes have been measured using a collimated reactor beam and BF3 counter, where the macroscopic cross section Σ(En) has been evaluated using the attenuation relation. The total macroscopic cross sections ΣT(En) and total attenuation coefficient µ(Eg) of neutrons and gamma rays have been calculated based on the database cross sections. Also, the total mass attenuation coefficient µeff(Eg)/ρ and µ(Eg)/ρ of gamma ray have been estimated and calculated using the measured results and XCOM code respectively. Measured and calculated results were compared and a reasonable agreement was found.

This work is concerned with the theoretical calculation of both the total mass attenuation coefficients ( μ / ρ ) for gamma rays and the effective removal cross-sections (Σ R ) for fast neutrons. Calculations were carried out for four types of concrete of different densities namely (1) dolomite–sand ρ =2.5 g.cm −3 , (2) barite–barite ρ =3.49 g.cm −3 , (3) magnetite–limonite ρ =3.6 g.cm −3 and (4) ilmenite–ilmenite ρ =3.69 g cm −3 . The total mass attenuation coefficients have been calculated at energies from 10 keV to 1 GeV using the XCOM (version 3.1) computer program and cross- section database for elements from Z =1 to 100. Also the effective removal cross-section have been calculated using the elemental composition of the concrete mixes. The calculated values are widely needed and used as a database for radiation shielding design of research reactors, power station and particle accelerators. The results were displayed in tables and figures. Calculated results were compared with those previously measured and a reasonable agreement was found between them.

The aim of this work is the evaluation and calculation of the heat generation and temperature rise in local ordinary concrete due to capture of thermal neutrons. The total thermal and reactor thermal neutron fluxes were measured, and the results were used to evaluate the heat generation and temperature rise in ordinary concrete. In addition, the computer code ANISN (VAX version) and neutron multigroup cross-section library EURLiB-4 were used to calculate the total thermal neutron fluxes and the heat generation and hence the temperature rise. The results were displayed in curves to show the distribution of thermal neutron fluxes and the heat generation as well as the temperature rise in the shield thickness. The results showed a good agreement between the measured and calculated values. The results showed also that the heat generation and the temperature rise have their maximum values in the first layers of the shield thickness.

The shielding effect of an iron sphere assembly has been tested for a Pu-α-Be neutron source placed in the center of the shield assembly. Emergent neutron and gamma spectra were measured with a stilbene scintillation counter. Discrimination between neutrons and gammas was achieved by the pulse shape discrimination technique based on the zero crossing method. Calculations have been made using the one-dimensional transport code ANISN-Westinghouse version (ANISN-W) and the EUR LiB 15/5 cross section data set. The agreement between measurements and calculations indicates that the cross section set and the calculation model are suitable for studying the iron shielding experiments over the neutron energy range 1.35–10 MeV and the gamma energy range 0.3–6 MeV. Total macroscopic cross sections for fast neutrons, linear attenuation coefficients for gamma rays and half-value thicknesses for neutrons and gammas for the whole energy range and at different energies have been obtained.

Measurements have been carried out to determine the spatial and energy distributions of fast neutrons, total gamma rays and secondary gamma rays in aluminum as a reactor wide structural material. Experimental work was performed using one of the horizontal channels of the ET-RR-1 reactor. For secondary gamma ray investigations, measurements were carried out for bare, cadmium filtered and boron carbide filtered reactor beams. Aluminum cylinders with a diameter of 100 mm and of different thicknesses have been used. Fast neutrons and gamma rays were measured using a neutron-gamma spectrometer with a stilbene scintillator. Discrimination against undesired pulses of neutrons and gamma rays was achieved by a pulse shape discrimination based on zero cross over technique. The measured pulse amplitude distributions of fast neutrons or gamma rays were transformed to neutron or gamma ray energy distributions using the differentiation method and the matrix correction method, respectively. The obtained data were displayed in the form of spatial fluxes and energy distributions and attenuation curves. Total fast neutron macroscopic cross sections, gamma rays linear attenuation coefficients and the relaxation lengths were obtained for the whole energy range and at different energies.

This work is concerned with the use of ilmenite concrete (density = 3.5 g/cm 3 ) prepared from local materials as a radiation shield. Comparison with a previously investigated ilmenite concrete (density = 4.6 g/cm 3 ) including iron punching additions in its constituents have been carried out as this type of concrete was chosen for the construction of the biological shield for the second experimental reactor in Egypt. The study reveals that the ilmenite concrete under investigation is more effective for the attenuation of fast neutrons. In addition the production of the hard capture secondary gamma rays have been greatly decreased by eliminating the iron punchings from the constituents of the ilmenite concrete under investigation which is also preferable from the economical point of view. Total macroscopic cross section for fast neutrons, linear attenuation coefficients for gamma rays and the relaxation lengths both for the whole energy range and at different energies have been obtained and comparison with other published data has been done when it is available. Two empirical formulae have been derived which calculate the total flux of fast neutrons and gamma rays for different types of concrete at different thicknesses. Good agreement between measured and calculated values using the two formulae has been obtained.

This work is concerned with the study of the attenuation properties of hematite-serpentine and ilmenite-limonite concretes for fast neutrons and γ-rays. Investigations were carried out using a collimated reactor beam of neutrons and γ-rays. The neutrons and γ-spectra have been measured by a scintillation spectrometer with stilbene crystal. Discrimination against undesired pulses of γ-rays or neutrons was achieved by a zero crossover technique. The obtained data are displayed in the form of spectra and attenuation curves. The obtained results were also used to derive the total cross sections for fast neutrons and the linear attenuation coefficients for γ-rays. Two semiempirical formulae have been derived to calculate the integral flux for fast neutrons and γ-rays behind different thicknesses of any of the two types of concretes under investigations using the corresponding integral flux obtained for the other type. A good agreement between the calculated and the measured values was obtained.

Composites of ethylene-propylene diene rubber and low density polyethylene were prepared with two different concentrations of boron carbide powder, namely, 47 and 57 wt%. The composites were investigated for their gamma and slow neutron radiation shielding properties. In addition, the temperature dependence of the electrical and thermal properties of the composites in the temperature range 30–150 °C, has been studied. The initial 1.5 cm thickness of the composite sample containing 57 wt% boron carbide sharply reduced the initial direct slow neutron flux by about 85%. The total macroscopic cross-section of this sample is 0.215 cm−1.

In this work a comparative study of the effect of iron loaded with B4C on the suppression of gamma-rays in a laminated shield of water, iron and either ordinary or limenite concrete have been carried out. It was found that B4C loaded with iron has a remarkable effect on the suppression of gamma doses all through the laminated shield. The suppression is more pronounced in the concrete layer. It was also found that the decrease of gamma doses in ilmenite concrete is ∼ 1.7 that one for ordinary concrete. Untersuchung des Einflusses von mit B4C angereichertem Eisen auf die Gesamt-Gamma-Dosis aus einem laminierten Schild Der Effekt von mit B4C angereichertem Eisen auf die Abschirmung von Gamma-Strahlen durch einen laminierten Schild aus Wasser, Eisen und entweder gewohnlichem oder Ilmenitebeton wird vergleichend untersucht. Es erweist sich ein bemerkenswerter Effekt des angereicherten Eisens auf die Abschwachung der Gamma-Dosis im gesamten Schild. Sie ist in der Betonschicht am ausgepragtesten. Die Abnahme der Gamma-Dosis in Ilmenitebeton ist etwa 1,7fach der in gewohnlichem Beton.

Measurements were carried-out to compare the attenuation properties of homogeneous shields and shields of two layers and three layers for fast neutrons and total gamma rays. These were performed by measuring the fast neutron and total gamma-ray spectra behind homogeneous shields of magnetite–limonite, ilmenite–ilmenite and magnetite–magnetite concretes. The two layers assembly consists of iron and one of the above mentioned concretes, while the three layers shield consists of water, iron and one of the previously mentioned concretes. All measurements were carried-out using a neutron-gamma spectrometer with stilbene scintillator coupled to a fast photomultiplier tube. Separation between pulses of recoil protons and recoil electrons was achieved by a pulse shape discrimination technique. The obtained data are presented in the form of neutron and gamma spectra measured behind homogeneous shields and shields of two layers and three layers. The spectra measured behind homogeneous shields were used to derive the total macroscopic cross-sections for neutrons and the total attenuation coefficients for total gamma rays.

This paper is concerned with the study of the leakage spectra of total gamma rays and fast neutrons from two types of heavy concretes made from local magnetite ores. The two types are the steel-magnetite concrete (ρ=5.11 g/cm3) and the basalt-magnetite concrete (ρ=3.05 g/cm3). The radiation source used was a collimated beam emitted from one of the horizontal channels of the ET-RR-1 reactor. Total gamma rays and fast neutrons were measured using a gamma-neutron spectrometer with a stilbene scintillator. Gamma ray or neutron events were rejected by a pulse shape discriminator based on the zero cross over method. The obtained data were displayed in the form of spatial fluxes and energy distributions and attenuation curves. The linear attenuation coefficients for total gamma rays, macroscopic removal cross sections for fast neutrons and half-value layers for gamma rays and neutrons have been obtained for the whole energy range and at different energies. Comparison with other published data has been done when such data are available. For comparison, the steel-magnetite concrete is a much better radiation attenuator than the magnetite and basalt-magnetite concretes, especially for thick shields.

The experimental measurements of the transmitted and the reflected (albedo) components of the capture secondary gamma rays, arising from interaction of reactor neutrons with heat resistant concrete shields have been carried out. This new type of concrete has been prepared from Egyptian ilmenite-limonite ores and has a density 3.44 gm/cm3. The measurements were performed using a collimated beam of reactor neutrons emitted from one of the horizontal channels of the ET-RR-1 reactor.