Thermal Neutron Attenuation Research Articles

Neutron radiography with a one-dimensional image sensor having a high linearity range

By a combination of a linear image sensor and a honeycomb collimator, neutron radiography imaging system with wide dynamic range was constructed. Using this system, the effective thermal neutron attenuation coefficients for several materials were evaluated systematically. The imaging system was also used to perform neutron computed tomography as well as to evaluate the CT-value for several metallic samples. The obtained CT values agreed well with the effective neutron attenuation coefficients.

Attenuation of thermal neutrons by an imperfect single crystal

A semi-empirical formula is given which allows one to calculate the total thermal cross section of an imperfect single crystal as a function of crystal constants, temperature and neutron energy E, in the energy range between 3 meV and 10 eV. The formula also includes the contribution of the parasitic Bragg scattering to the total cross section that takes into account the crystal mosaic spread value and its orientation with respect to the neutron beam direction. A computer program (ISCANF) was developed to calculate the total attenuation of neutrons using the proposed formula. The ISCANF program was applied to investigate the neutron attenuation through a copper single crystal. The calculated values of the neutron transmission through the imperfect copper single crystal were fitted to the measured ones in the energy range 3 - 40 meV at different crystal orientations. The result of fitting shows that use of the computer program ISCANF allows one to predict the behaviour of the total cross section of an imperfect copper single crystal for the whole energy range.

Radial attenuation of thermal neutrons around a joint in a straight duct in an ilmenite concrete shield

The thermal neutron flux has been measured radially around a duct consisting of two parts of different radii pierced in the center of an ilmenite concrete assembly. The concrete assembly has dimensions of 120 × 120 × 120 cm 3 and was arranged in front of one of the horizontal channels of the ET-RR-1 reactor. The presence of this air filled duct in the ilmenite concrete shield causes a relative increase in the neutron flux compared with that for solid ilmenite concrete. The difference in radii between the two parts of the air-filled duct causes a peak in the thermal neutron flux at a distance ≈10 cm from the second part of the duct which has the smaller radius. This increase in thermal neutron flux becomes more as the difference between the radii of the two parts of the duct increases It was found that the duct effect K E φ ducted φ solid increases with an increase in the difference between the radii of the two parts of the duct. An empirical formula has been derived to describe the radial attenuation of thermal neutrons. It can be fitted to the measured data within a factor (0.3–1.3).

Study of Filtration of Reactor Beam of Neutrons with Cadmium in a Multilayer Shield Containing Boron Carbide

Experimental measurements have been carried out to study the effect of cadmium on the distribution and attenuation of reactor thermal neutrons emitted from a reactor core and the new thermal neutrons produced in a heterogeneous shield of water, iron, iron + B4C and ordinary concrete. Measurements were performed using a reactor beam of neutrons filtered with cadmium emitted from one of the horizontal channels of ET-RR-1. It was found that the presence of cadmium sheet at channel exit causes a marked decrease in the thickness of the shield required to attenuate the thermal neutron flux by a certain factor.

Tolerance of canine brain to boron neutron capture therapy.

The normal canine brain tolerance to boron neutron capture therapy (BNCT), with which selective radiotherapy of the brain tumor is expected, was demonstrated by two different series of the irradiation experiment. Normal dogs were irradiated on the skull with thermal neutrons from the reactor after intravenous infusion of 10B-enriched Na2B12H11SH, and the intravascular dose at the brain surface was calculated as the radiation dose. In series 1, the brains of 3 dogs were irradiated 2000 to 3000 rads by the current therapeutic regimen of BNCT and were autopsied after three years observation. In series 2, 6 dogs were given various high doses up to 9000 rads to their brain by BNCT, and autopsied after 1 month and 1 year respectively. Tolerance was excellent in all the dogs, with no unusual findings clinical, neurological, or in laboratory data during the observation period, and no pathological changes by autopsy. When the reduction of the endothelial dose to one third of the intravascular dose and the quick attenuation of thermal neutrons are taken into consideration, the actual doses to the vascular wall in both experiments might be low enough to be tolerated by the normal brain tissue.

Attenuation of reactor thermal neutrons in a bulk shield of ordinary concrete

This work is concerned with the study of the distribution and attenuation of doses of thermal neutrons emitted directly from the core of 235U research reactor in ordinary concrete shields. In practice it is not possible to identify the reactor thermal neutrons in the emitted continous neutron spectrum, therefore, measurements were carried out by using a direct and cadmium filtered beam of reactor neutrons. All measurements were performed using Li 2B 4O 7:Mn thermoluminescent dosimeters. The data obtained were analysed and the dose distributions of reactor thermal neutrons were evaluated. A group of isodose curves were constructed which give directly the shape and thickness of the shield required to attenuate the intensity of doses of reactor thermal neutrons to specific values. In addition, the thermal neutron relaxation lengths in ordinary concrete were derived for disc collimated beam and infinite plane monodirectional sources.

In situ Root Studies Using Neutron Radiography1

AbstractMost studies of plant roots growing in soil involve destructive sampling or involve estimates of root growth only at a soil‐viewing surface interface where growth conditions are different from those in bulk soil. This research tested the feasibility of using preferential attenuation of thermal neutrons by roots to overcome these two disadvantages of current methods. Radiographs were obtained of soybean [Glycine max (L.) Merr.] and corn (Zea mays L.) roots growing through bulk samples of unsaturated soil.Plants were grown at 22 C in loamy sand soil wetted to 9% volumetric water content. The soil was contained in either 2.5‐ or 5.0‐cm thick aluminum boxes. At irregular intervals, the sample containers were transferred into a 6° diverging 0.04eV (thermal) energy neutron beam obtained from the Ames Laboratory Research Reactor, U. S. Energy Research and Development Administration, Iowa State University, Ames, Iowa. The samples were exposed to the neutron beam for 8 to 10 min. This exposure allowed about 5 × 109 neutrons/cm2 to strike an indium collector plate attached to the rear of the container. After the collector plate was removed from the neutron beam, photographic film was attached to the collector plate for 1 ½ hours; then the film was processed.The preferential neutron scattering by the roots allowed elongation rates of soybean radicles or seminal roots of corn to be determined easily through either 2.5‐ or 5‐cm thick soil samples. Small lateral roots of either species were less clearly distinguishable and resolution must be improved before elongation rates can be measured for laterals with a diameter smaller than 0.33 mm. Plants were not harmed visibly by the thermal neutron fluxes used in these experiments.

Long range effects of neutron capture therapy of cancer in mice II. Dose effect

An index is established which compares the outcome of neutron therapy experiments when only a single exposure is given to a transplantable malignant neoplasm in mice if the thermal neutron attenuation in passing through the tumor and tumor boron concentration at the time of exposure are known. A high degree of correlation is shown for the EXIT RAD INDEX in tables and graphs recording and summarizing studies done on some 3500 mice treated for thigh-implanted malignant neoplasms varying from 8–17 mm in dia. at treatment time. The end point was permanent, total regression of the neoplasm occurring without scarring Each animal with a treated neoplasm was followed until death. It is shown that this INDEX makes possible valid comparisons of procedures carried out over several years and in which exposure times vary from 0.1 sec to 23 min in three different reactors at two different boron dosages.

Reactor Flux Measurements Using Thermoluminescent Dosimetry

Thermoluminescent dosimetry (TLD) was successfully evaluated as an in-core thermal-neutron-flux determinant. The LiF crystals enriched with either /sup 6/Li or /sup 7/Li provided two effective neutron-gamma discrimination techniques. The first method used both types of crystals. The /sup 6/LiF dosimeters, which have large thermal-neutron cross sections, detected both neutrons and gamma radiation, while the /sup 7/LiF dosimeters, possessing negligible thermal-neutron attenuation characteristics, monitored the gamma component only. The dosimeters were inserted into a reactor for a known time interval and read on a commercially available detection system, and the difference in dosimeter exposure yielded a direct measure of neutron flux. The second technique used bare and cadmium-covered /sup 7/LiF dosimeters. The bare crystals detected reactor gammas, while those encapsulated in cadmium measured reactor gammas plus capture gammas from the Cd(n,..gamma..) reaction. The difference in exposures provided the capture-gamma contribution, which was proportional to reactor flux. Experiments using a subcritical and a TRIGA reactor revealed exposure rate to neutron flux sensitivities of 1.4 x 10/sup -7/ R/sec per phi and 2.6 x 10/sup -8/ R/sec per phi for the respective techniques. Accurate flux measurements were obtained over a range spanning 10/sup 2/ to 10/sup 12/ n/(cm/sup 2/ sec). At higher fluxes, more » the dosimeters experienced radiation damage and readings became unreliable. The TLD results were compared against BF/sub 3/ detection, foil activation, and fission chambers to derive an empirical exposure rate to the flux conversion factor. « less

On the determination by neutron radiography of the thickness of the carbonated layer of concrete based upon changes in water content

A description is given of the application of neutron radiography to the detection of water changes in concrete. As the attenuation of thermal neutrons in concrete is mainly attributable to interactions with hydrogen, even small changes in water content, such as those associated with carbonation, are observable. The thickness of the carbonated layer at the ends of the prisms of cross-sections of 25 mm × 25 mm was obtained from the radiographs as a region of lower water content. The water content is determinable to a relative accuracy of about ± 1.6%, and the spatial resolution of changes in water content amounts to some tenths of a mm.

Attenuation of Thermal Neutrons by Phonons in a Single Crystal. I

The effect of thermal vibration on the transmission or reflection of neutrons in a single crystal was investigated theoretically. The formula of imaginary potential which is used to get the absorption coefficient was given as a result of thermal inelastic scattering of neutrons. It was made clear that the surface integral taken over the scattering surface is necessary to obtain the imaginary potential of a single crystal for one-phonon inelastic scattering of slow neutrons.

軽水中の鉄層による中性子の減衰

The thermal neutron attenuation through iron layer bounded by light water was measured in the HTR experimental pool for six cases in which the iron layers had thickness ranging from 2 cm to 20 cm.In order to obtain the removal cross section of iron from measurements of neutron flux with and without iron layer, the following corrections were made.(1) The geometrical correction of the distance between source and shielding plate to account for the use of reactor core instead of fission plate as neutron source.(2) The correction for the effect of the difference between incident neutron spectrum on iron layer and fission neutron spectrum, obtained from the differrence of mean energy between these spectra and the energy dependence of the total cross section of iron.With these corrections, the removal cross section of iron was determined to be Σr=0.165±0.006 cm-1.This method should be applicable to other cases where a reactor is similarly used for the determination of removal cross section.The measured cross section of iron layers thinner than 10 cm was about 6% larger than similar layers of 1520 cm.

A method for prediction of the neutron attenuation in annular ducts☆

A calculational procedure for prediction of the attenuation of thermal and fast neutrons in straight air-filled annular ducts is described. The underlying theory, not presented here, takes into consideration arbitrary angular distributions of the source flux and treats both collided and uncollided contributions to the air gap flux. The procedure works also at short distances from the duct mouth.

A method for prediction of the neutron attenuation in annular ducts

A calculational procedure for prediction of the attenuation of thermal and fast neutrons in straight air-filled annular ducts is described. The underlying theory, not presented here, takes into consideration arbitrary angular distributions of the source flux and treats both collided and uncollided contributions to the air gap flux. The procedure works also at short distances from the duct mouth.

A comparative method for testing graphite used in nuclear reactors

The purity of the graphite used in graphite-moderated thermal reactors is a factor of prime importance. This article describes a method for testing the reactor feasibility of graphite. The gist of the method involves comparison of flux attenuation of thermal neutrons in the test graphite and in a standard graphite specimen. The method serves in determining the diffusion length L or the effective absorption cross section σ of graphite, where L0 and σ0 for the standard graphite are known. The error incurred in determining σ does not exceed 2.5%. The simplicity of the method recommends its use for large-scale testing of graphite under factory conditions.