Total Macroscopic Cross Section Research Articles

Identification of an Unknown Material in a Radiation Shield Using the Schwinger Inverse Method

The Schwinger method for solving inverse gamma-ray transport problems was proposed in a previous paper. The method is iterative and requires a set of uncoupled forward and adjoint transport calculations in each iteration. In this paper, the Schwinger inverse method is applied to the problem of identifying an unknown material in a radiation shield by calculating its total macroscopic photon cross sections. The gamma source (its energy and spatial distribution as well as the composition of the material) is known and the total (angle-independent) gamma leakage is measured. In numerical one-dimensional spherical and slab test problems, the Schwinger inverse method successfully calculated the photon cross sections of an unknown material. Material identification was successfully achieved by comparing the calculated cross sections with those in a precomputed material cross-section library, although there was some ambiguity when realistic measurements were used. The Schwinger inverse method compared very favorably with the standard single-energy transmission technique.

Analysis of neutron attenuation in boron-alloyed stainless steel with neutron radiography and JEN-3 gauge

This paper concerns the neutron attenuation behavior of boron-alloyed stainless steels, and quantitative estimations of secondary effects on the neutron transmission measurements, such as beam hardening, background and micro-structure of neutron absorber causing a decrease in the total macroscopic cross-sections. Systematic thermal neutron transmission measurements of boron-alloyed steel plates have been performed at different neutron radiography facilities and at a portable gauge called JEN-3, which is a practical neutron transmission setup for industrial demands in non-destructive material inspection. In addition, MCNP (Monte Carlo Neutron Particle transport) simulations were performed for a better understanding of the experimental results, the secondary factors affecting the total macroscopic cross-sections, and the characteristics of the JEN-3 gauge. The detectable thickness limit for 1.88wt% natural boron-alloyed steel plates was determined as 1–1.5cm depending on the contributions of the secondary factors in each facility.

Recognition of internal structure of unknown objects with simultaneous neutron and gamma radiography

Generally speaking in customary industrial and medical radiography, there is no tendency to reveal the nature of the samples. Ordinarily, the main objective of taking a radiograph is to show the position and dimension of unknown parts, inside the test object and to determine cracks, defects, etc. Whereas in radiography many important factors such as material cross-sections and build-up factors are also involved. In this paper, by using both neutron and gamma radiography techniques, some mathematical relations were successfully generated, in order to calculate the neutron and gamma total macroscopic cross-sections of some unknown elements in the presence of the other elements. For this work, some test pieces were defined and made of lead, silver, copper, Nickel, tin, graphite and polyethylene. The neutron radiography facility at Tehran Research Reactor (TRR) was used as mixed neutron and gamma radiography source (Proceedings of the Second World Conference on Neutron Radiography, Paris, France, pp. 25–32). On testing of a correction of the above-mentioned generated relations, a new technique of simultaneous neutron and gamma radiography was also investigated.

Recognition of internal structure of unknown objects with simultaneous neutron and gamma radiography

Generally speaking in customary industrial and medical radiography, there is no tendency to reveal the nature of the samples. Ordinarily, the main objective of taking a radiograph is to show the position and dimension of unknown parts, inside the test object and to determine cracks, defects, etc. Whereas in radiography many important factors such as material cross-sections and build-up factors are also involved. In this paper, by using both neutron and gamma radiography techniques, some mathematical relations were successfully generated, in order to calculate the neutron and gamma total macroscopic cross-sections of some unknown elements in the presence of the other elements. For this work, some test pieces were defined and made of lead, silver, copper, Nickel, tin, graphite and polyethylene. The neutron radiography facility at Tehran Research Reactor (TRR) was used as mixed neutron and gamma radiography source (Proceedings of the Second World Conference on Neutron Radiography, Paris, France, pp. 25–32). On testing of a correction of the above-mentioned generated relations, a new technique of simultaneous neutron and gamma radiography was also investigated.

Investigation of Radiation Attenuation Properties for Baryte Concrete

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 g·cm-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.

Influence of magnetite and boron carbide on radiation attenuation of cement–fiber/composite

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.

Natural rubber composites as thermal neutron radiation shields: II — H 3BO 3/NR composites

Different amounts of boric acid (H 3BO 3) were mixed with a conductive natural rubber (loaded with 40 phr of HAF carbon black) to get thermal neutron radiation shielding composites. It was found that the total macroscopic cross-section reaches 0.29 cm −1 at 30 phr of H 3BO 3. The dependence of thermal and electrical properties of such composites on the concentration of H 3BO 3 was also studied. Thermal oxidative aging was found to markedly affect the above properties.

Homogeneous and Multilayered Shields for Neutrons and Gamma-Rays

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.

Study for etm cross section measurements

An effective total macroscopic (ETM) cross section was proposed as a useful parameter to describe a transmission character of various materials for a neutron beam having a continuum energy spectrum. The ETM cross sections of several materials,determined as a function of effective energies, were measured using a beam quality indicator. Appreciable discrepancies of ETM cross sections were observed in comparison with total macroscopic cross sections measured by a time-of-flight technique. The reliability of ETM cross sections of Ti and Pb are reevaluated considering effects of scattering and gamma ray components.

Attenuation of Neutrons and Total Gamma Rays in Two Layers Shields

Experimental measurements have been carried out to study attenuation of neutrons and total gamma rays in two layer shields of made steel + magnetite-limonite, steel + ilmenite, limonite steel + water. Measurements were performed using a collimated beam of reactor neutrons and gamma rays emitted from one of the horizontal channels of T-R reactor. The transmitted neutron and gamma aspectra were measured by a neutrons and gamma spectrometer with NE-213 scintillator. Discrimination between neutrons and gamma pulses was achieved by a discrimination method based on pulse shape discrimination technique.The obtained results are displayed in the form of neutron and gamma ray spectra and their attenuation relations which are used to derive the total macroscopic cross sections for neutrons and total linear attenuation coefficient for gamma-rays.

A correlated study between effective total macroscopic cross sections and effective energies for neutron beams with continuous spectra

Two practically useful quantities have been introduced to characterize a continuous-energy-spectrum neutron beam and to describe transmission phenomena of the beam in the field of quantitative neutron radiography. These quantities are the effective energy instead of a peak energy or a mean energy of the spectrum and an effective total macroscopic (ETM) cross section instead of a total macroscopic (TM) cross section as defined for a monochromatic energy. Four neutron beams have been used to measure ETM cross sections at effective energies of 29.8, 17.2, 9.8 meV, and at the In resonance energy of 1.46 eV. Results are studied as a function of estimated effective energy, where the effective energy was estimated by a beam quality indicator (BQI) which has been proposed recently. Validity of ETM cross sections as a function of the effective energy is discussed and correlated with recent nuclear data.

Study for a novel tomography technique using an imaging plate

A sinogram is directly recorded onto an imaging plate for neutron tomography, which has excellent capability to detect neutrons. A sinogram-recording-system and outline of the data processing are explained briefly. A tomogram of a test object is reconstructed and shown as the preliminary result. The linearity of CT values was experimentally confirmed up to 1.6 cm −1 of effective total macroscopic cross section, but about 10% lower CT value was observed at 2.7 cm −1. The linearity limitation and statistical character of CT values are discussed with regard to the dynamic range of the system as well as the effect of background components.

Development of high-frame-rate neutron radiography and quantitative measurement method for multiphase flow research

In relation to the establishment of thermal neutron radiography as a measurement method with high accuracy and reliability, this paper reviewed the present status on the development of high-frame-rate neutron radiography with a steady thermal neutron beam and its application to multiphase flow researches. This review included also the present progress on the quantification of neutron radiographic image at Kyoto University, i.e. (1) quantitative method to measure void fraction of two-phase flow with thermal neutron radiography (Σ-scaling method), (2) influence of scattered neutrons on void fraction measured by neutron radiography, (3) measurement error of neutrons in a low neutron flux field, (4) error in void fraction measurement due to low gray level, and (5) measurement error due to low imaging speed Moreover, a new experimental approach on a total macroscopic cross section for thermal neutrons measurement by neutron radiography was presented. This paper revealed neutron radiography to be a promising visualization and measurement method in thermal hydraulic research.

Xenon poisoning method for the determination of the average thermal neutron flux, macroscopic fission and total absorption cross sections / Die Xenon-Vergiftungsmethode zur Bestimmung des durchschnittlichen thermischen Neutronenflusses, des makroskopischen Spaltungsquerschnittes und des totalen Absorptionsquerschnittes

The paper presents the xenon poisoning method as an experimental way to determine the average thermal neutron flux (Φ), the average macroscopic fission cross section (Σ f ) and the average total macroscopic absorption cross section (Σ a ) together with the reactivity change due to the 135 Xe poisoning. The predictions for the reactivity change due to the 135 Xe poisoning are in good agreement with the results of the experiments. The results of Φ, Σ f and Σ a obtained by using the xenon poisoning method are also in satisfactorily agreement with the predictions of TRIGAP and WIMS-D/4 computer codes. The temperature coefficients of the reactivity for the fuel and tank water have also been determined experimentally during this study.

Radiation attenuation and nuclear properties of high density concrete made with steel aggregates

The fast neutron and gamma ray spectra measured behind different thickness of steel scrap concrete with density of 4 g/cm3 have been studied. The mix proportions by weight of this type of concrete were 1 cement: 6.89 steel scrap: 2.9 sand and 0.5 Water. Comparison with a standard ordinary concrete of density 2.3 g/cm3 have been carried out. The measurements were made using a collimated beam of both gamma rays and neutrons emitted from one of the horizontal channel of the Egyptian Research Reactor-1. A fast neutron and gamma ray spectrometer with a stilbene crystal was used to measure the spectra of fast neutrons and gamma rays. Pulse shape discrimination using the zero cross over technique was used to separate the photon pulses from the electron pulses. The equation due to Schmidt has been modified and applied for determining the neutron effective removal cross sections (∼R) for steel scrap, ordinary, hematite–serpentine, ilmenite–limonite and ilmenite concretes. This equation gives results which are in good agreement with the measured values. The derived empirical equation in a previous work to calculate the neutron integral flux behind different thicknesses of different types of concretes, gives good results for steel scrap concrete under investigation comparing with the corresponding experimental data. Total neutron macroscopic cross sections, linear attenuation coefficients for gamma rays and the half-value layers for both radiations at different energies have been obtained for steel scrap concrete and comparing with the corresponding values of ordinary concrete. The results show that steel scrap concrete is better than ordinary, hematite–serpentine, ilmenite–limonite and ilmenite concretes from the radiation shielding point of view.

Use of the EUR LiB 15/5 data set for radiation shielding calculations in iron

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.

The Derivation and Proper Use of Stewart’s Formula for Thermal Neutron Self-Shielding in Scattering Media

The derivation and proper use of a formula originally published in 1959 for thermal neutron self-shielding computations in scattering bodies are presented. It is shown that the f{sub 0} entity in this formula must be computed using the total macroscopic cross section {Sigma}{sub t} instead of the macroscopic absorption cross section {Sigma}{sub a}, as was stated in previous publications. The validity of the formula is verified by Monte Carlo computations.

Quantitative Method to Measure Void Fraction of Two-Phase Flow Using Electronic Imaging with Neutrons

A quantitative method of image processing coupled with the neutron radiography technique is proposed to accurately measure the void fraction of a two-phase flow in a metallic duct. The spatial distribution of the dark current component is experimentally shown to be smooth, and the temporal variation cannot be ignored. Since the neutrons scattered in an object can be smoothed and reduced by setting the test section at a large distance from the converter, it is clarified that the corrections for the dark current and scattered neutrons can be represented by an offset. The offset value can be determined by using the total macroscopic cross section of the object ({Sigma}-scaling method). By comparing the calculated void fractions with the measured ones obtained by simulating the known void profile using a standard test section, the void fraction can be measured by this method within 2% error. The measurement error is estimated to be up to {approximately}10% when no correction for scattered neutrons is made or arbitrary offset values are used.

Investigation of radiations penetration in aluminum used in reactor technology

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.

Logarithmic transmittance and statistical characteristics on neutron computed tomography

A new quantity log-transminance will be defined as a parameter to evaluate neutron tomograms. Statistical characteristics of the log-transmittance are analyzed as functions of signal intensities taken into consideration the effects of scattered components. It is shown that the quantity should be valuable to evaluate an image quality related to a size and an effective total macroscopic cross section of an object on reconstructed CT values.