Measurement of moisture and hydrogen contents by intermediate neutron moderation

An intermediate neutron moderation method for measurement of moisture and/or hydrogen contents of small samples is presented. The sample is placed on the top face of a neutron howitzer, with a cadmium sheet between. Thermal neutrons resulting from intermediate neutron moderation in the sample are detected with a³He proportional counter placed on the sample, by a cadmium difference method. With a 500 mCi Am-Be neutron source, the limit of moisture detection for a 10×20×1.8 cm³ asbestos plate in 1 min count time is 0.5 wt.%. The precision of measuring the hydrogen contents of 250 ml hydrocarbons containing 112 mg H/ml is 0.9% under the same conditions.     

On-stream measurement of heavy water concentration by intermediate neutron moderation

A method of measuring heavy water concentration is described based on intermediate neutron moderation. With a 1.6 g²⁵²Cf neutron source and a BF₃ proportional counter, for 500 ml samples, accuracies of the measurements are within ±0.02 mol% in the concentration range of 0100 mol% with 210 minute counting. To evaluate the practical effectiveness of the method, a model apparatus was built for routine use in laboratory and on-stream measurements. It was successfully applied to observe the substitution process between light water and heavy water in an ion exchange resin tower at a nuclear power station.     

Determination of the hydrogen content in Nigeria palm oil by the neutron moderation method

The thermal neutron moderation analysis facility at the Institute of Experimental Physics, Debrecen, Hungary has been used to determine the weight percent of total hydrogen content in Nigerian palm oil. The facility utilizes the fast neutron moderation technique in which the intensity of reflected thermalized neutrons is proportional to the hydrogen content of the sample exposed to fast neutrons. Using a 100 cm³ sample the total hydrogen content in the oven-dry palm oil sample was found to be 12.0±0.1% within a measuring time of 5 minutes. The method is fast and can be used in plant quality control where the hydrogen content must be determined within specific limits.     

Determination of water in zeolites by a neutron reflection method

Water content in zeolites has been determined by an improved neutron reflection method using a Pu-Be neutron source of 10⁶ n·s^–1 intensity and a BF₃ counter. It was found that the water content of different types of zeolites collected in Hungary varies between 9 and 12 wt.%. The standard deviation of the determination does not exceed 0.5 wt.%. The matrix effect on the sensitivity and accuracy was also studied. An approximate relation is given between the count rate and the neutron physical parameters of the samples.     

Development of a Bonner sphere spectrometer with emphasis on decreasing the contribution of scattering by using a new designed shadow cone

In this study, the Bonner sphere spectrometer (BSS) used for measurement of neutron spectra based a BF₃ long counter. The most important problem for this system was a high count of scattered neutrons. This spectrometer was established by designed a new shadow cone with a smaller length and improved attenuation coefficient. Because of shortening the length of the shadow cone, distance of source and center of the sphere was decreased. Furthermore, external part of the thermal detector was covered with a suitable layer to reduce the contribution count of scattering neutrons. Experimental results show that BSS system with BF₃ long cylindrical counter, applying the proper developments, can be used in neutron fluence spectrometry.     

A review of intercomparison exercises run by NPL from 1989–1997

The flux of cold neutrons that is obtainable from various high energy netron sources is studied for a particular model of a cold neutron source when the cold moderation region of the apparatus is at 20, 70, and 298K. The maximum flux obtained with a californium-252 source was 2.7·10^?3 cold neutron per (cm²·second (source neutron)). This flux was obtained when the cold moderation region of the apparatus was at 20K and when the thermal moderator is either polyethylene or trimethylbenzene and the cold moderator is polyethylene. This flux should allow sensitive prompt and delayed neutron activation analysis measurements.     

A method for bulk hydrogen analysis based on transmission and back scattering of fast neutrons

A method was proposed for bulk hydrogen analysis. It is based on simultaneous detection of transmitted fast neutrons and back scattered thermal neutrons from the investigated samples by ³He detectors. The fast neutron beams were obtained from ²⁵²Cf and Pu–Be neutron sources. The experimental set-up as well as samples preparation were described. Incident thermal neutrons beams obtained from either ²⁵²Cf or Pu–Be sources, were used to investigate the samples by neutron backscattering. The results obtained from transmission and backscattering of fast neutrons were compared and discussed. The advantage and capabilities of the proposed method were presented. The results obtained using fast neutron beams are more sensitive than those obtained using thermal neutron beams. Validation procedures were proposed to credit the results.     

Neutron beam preparation with Am–Be source for analysis of biological samples with PGNAA method

Material analysis with prompt gamma neutron activation analysis (PGNAA) requires a proper geometrical arrangement for equipments in laboratory. Application of PGNAA in analysis of biological samples, due to small size of sample, needs attention to the dimension of neutron beam. In our work, neutron source has been made of ²⁴¹Am–Be type. Activity of ²⁴¹Am was 20 Ci which lead to neutron source strength of 4.4 × 10⁷ neutrons per second. Water has been considered as the basic shielding material for the neutron source. The effect of various concentration of boric acid in the reduction of intensity of fast and thermal components of the neutron beam and gamma ray has been investigated. Gamma ray is produced by (α, n) reaction in Am–Be source (4.483 MeV), neutron capture by hydrogen (2.224 MeV), and neutron capture by boron (0.483 MeV). Various types of neutron and gamma ray dosimeters have been employed including BF₃ and NE-213 detectors to detect fast and thermal neutrons. BGO scintillation detector has been used for gamma ray spectroscopy. It is shown that the gamma and neutron radiation dose due to direct beam is of the same magnitude as the dose due to radiation scattered in the laboratory ambient. It is concluded that 14 kg boric acid dissolved in 1,000 kg water is the optimum solution to surround the neutron source. The experimental results have been compared with Monte Carlo simulation.     

Reactor neutrons irradiation effect on the photocatalytic activity of SnO<Subscript>2</Subscript> thin films

SnO₂ thin films synthesized by sol-gel are irradiated with reactor neutrons up to fast neutron fluence of 9.6 × 10¹⁷ neutrons cm^–2 at 40°C. The influence of defects generated by neutrons irradiation, through the properties modification, on the photo-catalytic activity of SnO₂ films is investigated. It is found that the photoactivity of the irradiated films is enhanced after reactor neutrons irradiation. An improvement of 41% is observed for the sample irradiated at a neutron fluence of 9.6 × 10¹⁷ neutrons cm^–2. This is attributed to several parameters modified by the reactor neutron irradiation principally the crystallite size and space charge region which are closely related to the photocatalytic performance.     

Estimation of the possibilities of a neutron generator-based source of thermalized neutrons for activation analysis

Characteristics of a source of thermal neutrons based on an evacuated NG-400 neutron generator with the maximum flux (Φ_f) 2 × 10¹¹ neutron/s for 14 MeV neutrons and 2 × 10⁹ neutrons/s for 3 meV neutrons have been investigated. The possibilities of its application for neutron activation analysis have been estimated. The distribution, composition, and density (φ_T) values of the thermal neutron flux have been measured in the inner cavity of the moderator using activation detectors. φ_T was 2 × 10⁸ and 2 × 10⁶ neutrons/cm² s for thermalized neutrons with energies of 14 and 3 MeV, respectively. The possibilities of the apparatus have been estimated theoretically and experimentally for the cases of thermalized neutrons of 14 MeV and 3 MeV.     

K 0-prompt gamma ray activation analysis for estimation of boron and cadmium in aqueous solutions

K ₀-PGAA (prompt gamma activation analysis) has been used to estimate boron and cadmium contents in industrial and environmental materials from some different local areas as well as, samples from different locations in Burullus and Qarun Lakes. A high efficiency Compton background suppression gamma-ray spectrometry by anti-coincidence counting with a NaI(Tl) shield around a central HPGe detector for in beam PGAA using a ²⁵²Cf neutron source has been calibrated and described in this paper. The facility is principally designed for measurement of the prompt gamma-ray spectra obtained due to thermal neutron capture using 1951.14 keV gamma line of ³⁵Cl as the internal mono-standard comparator. A calibration curves were developed in which a set of boron and cadmium standards were tested and the count rate to boron and cadmium mass curves were determined. This set of boron and cadmium measurements was compared with a method for determining composition using K ₀-PGAA. Conventional prompt neutrons capture Gamma-ray results were in a good agreement with the data obtained by K ₀-PGAA method. Detection limits and self-shielding study are presented.     

Evaluation of an alternative convenient irradiation system for determination of emission rate of radio-isotopic neutron sources

In present work, an alternative irradiation system based on a symmetric cylindrical tank filled with a moderator containing hydrogen, which was equipped with a NaI(Tl) scintillation detector, was proposed for using in determination of neutron flux. This irradiation system was designed by MCNP4C code, with considering a ²⁴¹Am–Be neutron source in several volumes and different materials. When the neutron is captured by hydrogen, a 2.22 MeV prompt gamma-ray is emitted. The gamma pulse-height spectrum shows a photo-peak around 2.22 MeV whose net area is proportional to the total emission rate of neutron. The simulation result showed that a cylindrical tank with 110 cm diameter and height filled with water can be a suitable system for neutron source strength calibration. Furthermore, a proper two-layer shielding must be placed between the source and detector for preventing neutrons and gamma rays to directly enter the detector.     

Performance comparison of 2.8 MeV and <Superscript>241</Superscript>Am-Be neutrons based moisture measurement setups

Performance of a ²⁴¹Am-Be neutron source-based and 2.8 MeV neutrons-based moisture measurement setups have been compared using Monte Carlo simulation. In the setup fast neutrons transmitted through the sample were detected by a fast neutron detector, which was placed behind a massive long double truncated collimator. The setup geometry was optimized to detect maximum effect of 1–7 wt.% moisture on the neutron intensity transmitted through the sample. The yield of neutrons transmitted through concrete, coal, wood and soil samples containing 1–7 wt.% moisture was calculated for 2.8 MeV neutrons and neutrons from an ²⁴¹Am-Be source. The slopes of the fast neutron intensities transmitted through the samples vs. their moisture contents are very sensitive to the neutron energy and the sample composition. Higher slopes have been observed for the samples with larger bulk density. The slopes of fast neutron yield show dependence on the incident neutron energy. Larger slopes have been observed for neutrons with samller energy. Due to the overall large slopes of the transmitted intensity data of the samples for 2.8 MeV neutrons, it is expected to achieve better sensitivity in moisture measurements for a 2.8 MeV neutrons based moisture setup.     

Investigation of fast neutron shielding characteristics depending on boron percentages of MgB<Subscript>2</Subscript>, NaBH<Subscript>4</Subscript> and KBH<Subscript>4</Subscript>

The macroscopic cross-section Σ and average neutron fluence in matter Φ are usable factors to comment neutron shielding property of samples. In this paper, we have used MgB₂, NaBH₄ and KBH₄ samples including different percentages of boron. Neutron macroscopic cross-section measurements of them have been done by using a source of mono-energetic neutrons (E _eff = 4.5 MeV ²⁴¹Am–Be). Average neutron fluence values and double differential fast neutron flux distributions of each samples calculated by using FLUKA Monte Carlo code. Also half value layers (HVLs) of samples are compared to paraffin which is one of the most neutron moderators. As a result, growing boron concentration can raise neutron shielding property of materials.     

Search for neutrons from cold nuclear fusion

An attempt was made to detect neutrons from the so-called cold nuclear fusion of deuterium in palladium and titanium, both saturated with deuterium: the palladium electrolytically and the titanium from gas phase. The measurements were performed in a tunnel located 30 m deep in limestone, using³He filled proportional counters surrounded by water for neutron moderation. In all cases the detected neutron flux was practically equal to the background level. Very low upper limits to the neutron source strength were obtained from this experiment: 2×10^–4 n.s^–1g^–1 Pd and 4.3×10^–4 n.s^–1g^–1 Ti on the 1 level.     

Bestimmung von Cadmium,Bor und Indium in wäßrigen Lösungen durch Neutronenabsorption

Cadmium, boron and indium have high cross-sections different from other elements. Therefore they absorb slow neutrons strongly. Applying this property, it is possible to determine those elements rapidly and selectively. The arrangement consists of four parts, neutron source (Am 241-Be), moderator, cell for the aqueous solutions and BF₃-neutron counter. This method is suitable for the analytical laboratory and especially for continuous process control. The apparatus with the shielding is described exactly. Furthermore, the problem of the statistical error and detection limit is discussed.     

Measurement of neutron flux and albedo of water for thermal neutrons with foils of indium in a subcritical nuclear reactor

A subcritical nuclear reactor, Model 9000, Nuclear Chicago, is installed and operating at the Aristotle University of Thessaloniki, in the Atomic and Nuclear Physics Laboratory, at Thessaloniki, Northern Greece. The fuel is about 5500 lbs (2495 kgs) natural uranium metal (U₃U₈), the moderator about 3600 lbs (1633 kgs) light water, H₂O and the reflector is also light, water, H₂O. The lattice core is hexagonal, 42 inches (1.07 m) high and of 35 inches (88.90 cm) maximum diameter. The neutron source at the core is Am-Be 5 curies (185 GBq), 1.1·10⁷n·s⁻¹. The reactor is used for the activation of various materials by neutrons such as indium, the determination of the thermal neutrons flux, the horizontal and the vertical distribution of the neutron flux, material buckling, B, and geometric buckling, B, the parameters of the reactor, and the albedo of water for thermal neutrons with foils of indium. This revised version was published online in July 2006 with corrections to the Cover Date.     

Physical and chemical forms of radionuclides of 38S, 38Cl, 39Cl, and 82Br, produced at a high-energy proton accelerator facility

High-energy protons and neutrons produce various radionuclides in the air, mainly through the nuclear spallation of atmospheric elements. Air samples were collected from the EP-2 tunnel of the KEK 12 GeV proton synchrotron facility with a filter pack, which consisted of a membrane filter for aerosols, a Na₂CO₃-impregnated filter for acidic gases, and an activated carbon fiber filter for non-acidic gases. Sulfur-38 was found on the membrane and Na₂CO₃-impregnated filters, ³⁸Cl and ³⁹Cl were on the membrane, Na₂CO₃-impregnated, and ACF filters, and ⁸²Br was only on the ACF filter. The results on the relative abundances of aerosol and gaseous acidic components have indicated that ³⁸Cl produced by thermal neutron capture, which is the main reaction for ³⁸Cl production in the EP-2 tunnel air, are likely not to be present as aerosol or acidic gas. This was similar to the case of ⁸²Br produced by thermal neutron capture.     

Experimental and Monte Carlo simulation on new manganese bath facility for absolute neutron source emission rate measurement at ENEA-INMRI

This work deals with the absolute measurement of the neutron emission rate from a ²⁴¹Am–Be source by means of the manganese sulphate bath technique, which is the principal method for the absolute determination of the neutron emission rate from radionuclide neutron sources. The facility consists of a spherical container filled with an aqueous solution of manganese sulphate with a ²⁴¹Am–Be neutron source placed at the center. As well known, neutrons from the source, after having been thermalized by the aqueous solution, undergo neutron capture by hydrogen, manganese, sulphur, and oxygen nuclei, thus inducing a certain activity to the solution. Subsequent gamma spectrometry measurements of ⁵⁶Mn activity generated by ⁵⁵Mn neutron activation allows to determine the neutron emission rate of the source, The experimental activity has involved a variety of measurement techniques and calculation procedures, ranging from neutron reactor activation to liquid scintillation counting and Monte Carlo calculations. Neutron activations of ⁵⁵Mn samples has been carried out with the TRIGA reactor of the ENEA-Casaccia Research Centre, and ⁵⁶Mn activated samples were subsequently characterized by liquid scintillation counting, in order to obtain reference standards for the calibration of the NaI(Tl) scintillation detectors utilized to record gamma-ray emission from ⁵⁶Mn. Monte Carlo calculations, carried out by the MCNPX code, were required to calculate neutron transport within the sulphate manganese bath, in particular to determine ⁵⁵Mn neutron capture probability, and (n, α) and (n, p) concurrent reactions, as well as the neutron leakage. Such a procedure has allowed to maintaining the neutron emission rate uncertainty well below 1 %. All the measurements have been carried out at the ENEA-Casaccia Research Centre by the Italian National Institute of Ionizing Radiation Metrology.     

A transportable neutron radiography system

A transportable neutron radiography system, incorporating a 50 mg ²⁵²Cf source, has been simulated using the MCNPX code. The materials considered were compatible with the European Union Directive on ‘Restriction of Hazardous Substances’ (RoHS) 2002/95/EC, hence excluding the use of cadmium and lead. The design was optimized with respect to neutron moderation, shielding and collimation. High density polyethylene was chosen as the material for moderator and also shielding, which was further enhanced with layers of bismuth and borated polyethylene. Variable values for the collimator ratio were calculated. With suitable aperture and collimator design it was possible to optimize the neutron radiography parameters. Beam filters also were treated in order to improve the results. The proposed system has been considered with a wide range of radiography parameters, which are comparable with neutron radiography facilities from low power reactors.