Total Macroscopic Cross Section Research Articles

Measurement of fast neutrons and secondary gamma rays in graphite

The spatial fluxes and energy distributions of fast neutrons, total gamma rays and secondary gamma rays transmitted through different thicknesses of graphite have been measured. The graphite samples were arranged in front of one of the horizontal channels of the ET-RR-1 reactor. Gamma ray measurements were carried out for bare, cadmium filtered and boron carbide filtered reactor beams. A fast neutron and gamma ray spectrometer with a stilbene crystal was used to measure the spectrum of fast neutrons and gamma rays. Pulse shape discrimination using the zero cross over technique was used to distinguish the proton pulses from the electron pulses. The total fast neutrons macroscopic cross section and the linear attenuation coefficient for gamma rays were derived both for the whole energy range and at different energies. The obtained values were used to calculate the relaxation lengths for fast neutrons and gamma rays.

On the utilization of heavy concrete for radiation shielding

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.

Visualization and Measurements of Two-Phase Flows in Metallic Ducts Using Neutrons as Microscopic Probes. 3rd Report. Quantitative Measurement of Neutron Radiography Image.

The quantitative measurement of an image obtained by neutron radiography technique was proposed in order to accurately measure the void fraction of a two-phase flow in a metallic duct. It was shown experimentally that the spatial distribution of the dark-current component was homogeneous and the temporal variation could not be ignored. Since the scattered neutrons falling on the image converter could be homogenized by setting the test section at a distance from the converter, it was clarified that the corrections for dark-current and scattered neutrons could be represented by an offset value. It was proposed that the offset value could be determined by using the total macroscopic cross section of the material (Σ-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 could be measured by this method within 2% error. The measurement error was estimated to be up to 10% when no corrections for scattered neutrons were made or no arbitrary offset value was used.

Ethylene-propylene diene rubber/low density polyethylene/boron carbide composites as neutron shields

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.

Macroscopic Cross Section Measurements and Defect Detection in Materials Using Neutron Radiography Technique

Neutron radiography (NR) technique was applied to analyze behaviors of thermal neutrons in slabs, which have a wide range of the ratio r, of scattering to non-scattering cross sections. Transmittances were measured for various materials as basic data for NR study. Scattering components were estimated and subtracted from observed neutron intensities measured behind slabs. Effective total macroscopic cross sections were estimated and were found they agreed with given values of ASTM E748-80 within ±20%. Defects (voids and impurities) were also analyzed, and their detection limits for thickness t were estimated for materials of various r as a function of t.

An equivalence principle for the adjoint neutron flux in a lattice cell

The adjoint flux calculated by multigroup effective cross sections generated by considering the conservation of reaction rates in the direct flux has a noticeable error, especially in the resonance energy range. An equivalence principle for adjoint flux is developed for the calculation of multigroup effective cross sections to be used in adjoint flux calculation. These multigroup effective cross sections can be tabulated as functions of two independent parameters; namely the equivalent cross section and the total macroscopic cross section of the moderator medium. The numerical calculations show a linear dependence of the multigroup effective cross sections on the second parameter.

Macroscopic Cross Section Measurements and Defect Detection in Materials Using Neutron Radiography Technique.

Neutron radiography (NR) technique was applied to analyze behaviors of thermal neutrons in slabs, which have a wide range of the ratio r, of scattering to non-scattering cross sections. Transmittances were measured for various materials as basic data for NR study. Scattering components were estimated and subtracted from observed neutron intensities measured behind slabs. Effective total macroscopic cross sections were estimated and were found they agreed with given values of ASTM E748-80 within ±20%. Defects (voids and impurities) were also analyzed, and their detection limits for thickness t were estimated for materials of various r as a function of t.

Measurement of macroscopic and microscopic thermal neutron cross sections of V, Co, Cu, In, Dy and Au using neutron self-absorption properties

In this study, the total thermal neutron macroscopic and microscopic cross sections of V, Co, Cu, In, Dy and Au were measured using neutron self-absorption properties. Pure foils of these elements with various thicknesses were irradiated using a 5 Ci238Pu−Be source. After the irradiation, the gamma-spectra of their radionuclides were recorded by an HpGe detector. The gamma-photopeak areas of interest were determined by evaluating the gamma-spectra obtained from the foils. They were plotted as a function of foil thickness for each element. Then a non-linear least-squares fitting method was applied to the functions, and the total thermal neutron macroscopic and microscopic cross sections of V, Co, Cu, In, Dy and Au were obtained.

A simple model for neutron radiography of uranium-plutonium mixed oxide fuel pins

Neutron radiography has been used for monitoring plutonium enrichment in uranium-plutonium mixed oxide fuel pellets inside welded nuclear fuel pins by correlating the optical density of radiographs at the centre of the pellets and plutonium enrichment. Optical density data corresponding to different thickness of the pellets starting from the centre towards the periphery was generated by microdensitometer scanning of neutron radiographs of the experimental fuel pins. An attempt has been made to correlate the optical density at points corresponding to different thickness segments of the pellets and thermal neutron interaction probability (product of the total macroscopic neutron cross section and the distance traversed by the neutrons). Based on the experimental data generated, a simple model for transmission of neutrons through nuclear fuel pins has been evolved. Using this model, it is possible to predict the optical density of plutonium bearing fuel pins containing pellets of different composition or diameter in neutron radiographic investigations.

Effects of reflector on intensity of thermal neutrons emitted from moderator for pulsed neutron source.

Intensity of the thermal neutrons emitted from the moderator with a reflector was calculated to study the effects on the intensity caused by a macroscopic total neutron cross section and an average logarithmic energy loss of the reflector materials. A reflector with a large macroscopic total neutron cross section produced higher thermal neutron intensity than that with a small cross section if they had the same average logarithmic energy loss. Among the reflectors with the same total macroscopic neutron cross section, the thermal neutron intensity was not changed by decreasing the average logarithmic energy loss to a range less than about 0.1 but above this value the intensity was weakened. From this result it was found that a large macroscopic total cross section and a small average logarithmic energy loss are preferable characteristics for reflector materials. As actual reflector materials, three reflector materials were examined, namely beryllium, graphite and lead, which are now considered to be candi...

A Study of the Neutron Noise Transmission Characteristics of Nonmultiplying Media in Liquid-Metal Fast Breeder Reactors by Neutron Wave Propagation Technique

In this paper, an attempt has been made to investigate the noise transmission characteristics of nonmultiplying media of liquid-metal fast breeder reactors (LMFBRs) and study its implications on the detection of malfunctions in LMFBR cores by using out-of-core detectors and noise analysis methods. Neutron wave propagation technique has been used to study the problem by employing different approximations such as infinite and finite medium, one- and two-group diffusion theory, and multiregion and multigroup diffusion theory approximations. It has been found that reactor core noise will be transmitted to the out-of-core detectors with equal attenuation for all frequencies, ω < (ωΣt)min where Υ is the speed of neutrons and Σt is the total macroscopic removal cross section of the medium. For normal in-reactor vessel nonmultiplying media, (ΥΣt)min is of the order of 1 kHz. However, for materials like graphite if used as a moderator surrounding the out-of-core detectors, the limit (ΥΣt)min can be as low as 10 Hz. Reactor noise of malfunctions due to thermal events inside the reactor core such as sodium boiling lies in the frequency range of 2 to 15 Hz for integral boiling and goes up to 1 kHz for local boiling. Noise due to mechanical events is also a high frequency phenomenon. Therefore for detecting the malfunctions due to thermal and mechanical events in LMFBR cores by out-of-core detectors and noise analysis methods, one has to keep in mind that for moderating materials like graphite used in the surroundings of detectors, a band limited noise in reactor may be transmitted to detector locations in a distorted way and since high frequency noise is likely to be attenuated more, it will pose a problem in detecting the malfunction in its incipient stage.

Error analysis and convergence for the spectral synthesis method with interfaces

Error estimates and convergence are studied for the Galerkin-type spectral synthesis method relative to the continuous-energy, continuous-space, time-independent neutron diffusion equation. The diffusion coefficient and total macroscopic cross section are both space and energy dependent, and interfaces may be present. The estimate is in an energy type norm. The basic result shows that the error is optimal in the sense of being of the same order as the error provided by the best approximation to the actual solution in the approximation space where the spectral synthesis solution is found.

Sputtering Calculations with the Discrete-Ordinates Method

The purpose of this work is to investigate the applicability of the discrete-ordinates (S/sub N/) method to light ion sputtering problems. In particlar, the neutral particle discrete-ordinates computer code, ANISN, is used to calculate sputtering yields. No modifications to this code are necessary to treat charged particle transport. However, a cross-section processing code is needed for the generation of multigroup cross sections; these cross sections include a modification to the total macroscopic cross section to account for electronic interactions and small-scattering-angle elastic interactions. The discrete-ordinates approach enables calculation of the sputtering yield as functions of incident energy and angle and of many related quantities, such as ion reflection coefficients, angular and energy distributions of sputtering particles the behavior of beams penetrating thin foils, etc. The results of several sputtering problems as calculated with ANISN are presented.

Monte Carlo Sampling with Continuously Varying Cross Sections Along Flight Paths

A scheme is proposed to sample the distance to collision, without resorting to numeric integration, when the total macroscopic cross section varies in an arbitrary manner along projected particle flight paths. This sampling scheme is utilized in some example problems to illustrate methods of optimizing the sampling of the collision point in such a manner as to minimize relative errors for a fixed amount of computation time.

Diffusion Length for Arbitrarily Anisotropic Scattering

For the case of arbitrarily anisotropic scattering in monoenergetic neutron transport theory with no multiplication, the smallest root κ0 of the determinantal equation (which is equal to the inverse of the diffusion length) is considered as given by an infinite series in powers of the absorption parameter, where Σa is the macroscopic absorption cross section and Σ is the total macroscopic cross section. It is shown that αm depends only on the first m Legendre moments of the scattering probability.

Total Cross Sections of Hydrocarbons and Tissue

The total neutron cross sections of several organic materials are measured in the energy region 0.0I5 ev < E < 0.2 ev. The data are to be used in the application of the Neutron cross sections were studied for formic, acetic, propionic, and butyric acids; ethyl, propyl, and butyl alcohols; methyl-ethyl ketone; and human muscle, brain, and bone tissues. A graphical comparison is shown of the total macroscopic cross section of n-butyl alcohol and muscle tissue to that for water. (N.W.R.)