Prompt-gamma neutron activation analysis using an isotopic neutron source

This paper describes an experimental set up using an isotopic neutron source with graphite, preferably to a hydrogen-rich compound, as moderator. This low thermal flux source has been employed in PGNAA experiments. Only those elements having a high cross section can be quantitatively analyzed at low concentration levels, mainly, Gd, Cd, Sm, B and Hg; the other elements can only be probed at macroscopic levels. An application to the determination of boron content in collyria examplifies the possibilities of the technique.     

Correction for neutron self-shielding in large-sample prompt-gamma neutron activation analysis

When a sample is analysed with neutron activation analysis (NAA) neutron self-shielding and gamma self-absorption affect the accuracy. Both effects become even more important when the mass of a sample analysed is changed from small (say, 1 g) to large (say 30 kg). Therefore, corrections have to be carried out. In this article only the correction method for neutron self-shielding is considered for a thermal neutron beam irradiating large homogeneous samples for prompt-gamma NAA (PGNAA). The correction method depends on the macroscopic scattering and absorption cross sections of the sample. To avoid doing experiments with samples with different macroscopic scattering and absorption cross sections, the Monte Carlo model MCNP is applied in the development of the correction method. The computational development of the method to determine these cross sections through flux monitoring outside the sample is described.     

Optimization of Linac-based neutron source for thermal neutron activation analysis

Journal of Radioanalytical and Nuclear Chemistry - A 20 MeV electron Linac based neutron beam has been shaped for neutron activation analysis (NAA) technique. A beam shaping assembly (BSA)...     

Study of a neutron radiography system using Cf neutron source

This paper presents a study about moderation and collimation of a neutron radiography system using ²⁵²Cf. A Monte Carlo Code, MCNP4B, has been used to obtain a maximum and more homogeneous thermal neutron flux in the collimator outlet next to the image plane. Among the various moderator materials investigated, high density polyethylene proved to be the most efficient, with a thermalization factor of 56 cm². Using a collimator design assembly it was possible to obtain a normalized thermal neutron flux, at the image plane of 6×10⁻⁶ n cm⁻² s⁻¹ at an effective collimator ratio of 7.5, or 3.2×10⁻⁷ n cm⁻² s⁻¹ at an effective collimator ratio of 50.     

Indirect neutron absorptiometry

A preliminary account is given of the use of the neutron absorption of suitable elements (such as gadolinium) for the determination of non-absorbing elements (such as fluorine) by measuring the absorption of a dissolved precipitate (in the case quoted, gadolinium fluoride). This indirect determination by neutron absorption gives reproducible results rapidly and non-destructibly. The applications of the method are being studied.     

Monitoring of neutron flux changes in short-lived neutron activation analysis

It is well known variations in neutron fluxes can adversely affect the final result in neutron activation analysis. The monitoring of neutron flux changes are usually described for medium and long-lived NAA using foils of cobalt, gold, zirconium, etc. However, for short-lived neutron activation analysis there appears to be no systematic study of the variations of the neutron flux. With our new automatic pneumatic system, where irradiation timing, decay and counting and position are very reproducible, we have performed a series of experiments using thermal and epithermal neutrons using aluminum wire as a monitor to monitor the neutron fluxes. Our experiments confirm that neutron flux fluctuations in the worst case can be up to ±12 % with a SD of 2–3 %. This effect can be seen regardless of the irradiation time and must be taken into consideration to achieve the best result.     

Gamma-ray tables for neutron fast neutron and photon activation analysis

In the past several years, a number of gamma-ray tables have been published. All of the tables limit themselves to those nuclides formed in (n, γ) reactions or at most to a few of the more prominent (n, 2n) products. A number of investigators are developing activation analysis methods using fast neutrons or photons. These investigations have been hampered by the lack of tables of gamma-rays of those nuclides that can be formed by (n, 2n), (n, p), (n, α), (n, n′) and (n, d) reactions. Since (γ, n) yields the same product as (n, 2n) the table would also be useful to photon activation studies. The authors have compiled such a table. Section II: list of the gamma-rays for all isotopes in order of increasing energy will be published in the J. Radioanal. Chem., 25 (1975) No. 2.     

Spatial neutron flux distributions around A 14 MeV neutron generator

The neutron flux distribution in the vicinity of 30, 20 and 10 mm diameter targets is measured by irradiating concentric ring-type iron monitors at different distances from the target and counting the induced⁵⁶Mn activity. Considering the many uncertainties, satisfactory agreement was found between theory and experiment. Aspirant of the N.F.W.O.     

Fast neutron induced fission neutron spectra below the incident energy

A fast pneumatic transfer system for an instrumental neutron activation analysis and delayed neutron counting system were reconstructed with new designs of a functional improvement at the HANARO research reactor in 2006. The design, conception, operation and control of these systems are described. Also the experimental characteristic parameters by a functional operation test and an irradiation test of these systems, such as the transfer time, the neutron flux, the temperature of the irradiation position with an irradiation time, the radiation dose rate when the rabbit is returned, etc., are reported to provide a user information as well as for the management and safety of the reactor.     

Advanced compact accelerator neutron generator technology for active neutron interrogation field work

Due to a need for security screening instruments capable of detecting explosives and nuclear materials there is growing interest in neutron generator systems suitable for field use for applications broadly referred to as active neutron interrogation (ANI). Over the past two years Thermo Electron Corporation has developed a suite of different compact accelerator neutron generator products specifically designed for ANI field work to meet this demand. These systems incorporate hermetically-sealed particle accelerator tubes designed to produce fast neutrons using either the deuterium-deuterium (E _n = 2.5 MeV) or deuterium-tritium (E _n = 14.1 MeV) fusion reactions. Employing next-generation features including advanced sealed-tube accelerator designs, all-digital control electronics and innovative housing configurations these systems are suitable for many different uses. A compact system weighing less than 14 kg (MP 320) with a lifetime exceeding 1000 hours has been developed for portable applications. A system for fixed installations (P 325) has been developed with an operating life exceeding 4500 hours that incorporates specific serviceability features for permanent facilities with difficult-to-access shield blocks. For associated particle imaging (API) investigations a second-generation system (API 120) with an operating life of greater than 1000 hours has been developed for field use in which a high resolution fiber-optic imaging plate is specially configured to take advantage of a neutron point-source spot size of ∼2 mm.     

Instrumentation for neutron spectroscopy

The basic principles of operation of most types of neutron spectrometers used by chemists are reviewed. The merits and constraints of each type is explained in terms of energy range, resolution, data rate and momentum transfer range so that the reader can understand why particular experiments are, or are not, done.     

Continuous neutron activation analysis

A review on the uses of neutron activation techniques for on-stream analysis.     

A neutron activation analyzer

Summary Dubbed “Analyzer” because of its simplicity, a neutron activation analysis facility for short-lived isomeric transitions is based on a low-cost rabbit system and an adaptive digital filter which are controlled by a software performing irradiation control, loss-free gamma-spectrometry, spectra evaluation, nuclide identification and calculation of concentrations in a fully automatic flow of operations. Designed for TRIGA reactors and constructed from inexpensive plastic tubing and an aluminum in-core part, the rabbit system features samples of 5 ml and 10 ml with sample separation at 150 ms and 200 ms transport time or 25 ml samples without separation at a transport time of 300 ms. By automatically adapting shaping times to pulse intervals the preloaded digital filter gives best throughput at best resolution up to input counting rates of 10⁶ cps. Loss-free counting enables quantitative correction of counting losses of up to 99%. As a test of system reproducibility in sample separation geometry, K, Cl, Mn, Mg, Ca, Sc, and V have been determined in various reference materials at excellent agreement with consensus values.     

Multiparameter neutron activation analysis

An advanced neutron activation technique has been developed for the accurate analysis of elemental and isotopic fissile material required in nuclear safeguards, nuclear material standardization and other applications. It is based on reactor neutron flux spectrum differentiation by cadmium screening and multistandard calibration, including the solution of a second order equation system or of computerized calibration curve fitting, taking into account the thermal neutron flux depression. Some discrepancies at high enrichments have still to be eliminated in order to achieve the required measurement accuracy.     

Epithermal neutron activation analysis

Different modes of epithermal neutron activation analysis are described. The principles and examples are given for each.     

Monitoring of gamma emission and neutron transmission during boron neutron capture treatment delivery

The ability to map boron and hydrogen distributions in the body is paramount to the success of boron neutron capture therapy (BNCT). We investigated treatment-time quantitative mapping of these distributions by detecting (i) 0.48 MeV de-excitation photons from neutron capture by boron-10; (ii) 2.22 MeV photons from neutron capture by hydrogen; and (iii) transmitted neutrons. Monte Carlo simulations reported no detectable difference when ¹⁰B in tumour was varied from 0 to 50 ppm, and when the tumour size was varied from 0.0 to 9.5 cm³.     

A study of measured neutron elastic differential neutron cross sections for 23Na

Elastic neutron scattering angular distributions from ²³Na have been measured for incident neutron energies between 1.0 and 4.0 MeV at the University of Kentucky Accelerator Laboratory using neutron time-of-flight techniques for the scattered neutrons. This is an energy region in which existing data are very sparse. Measurements are compared with the predictions of the light particle-induced reaction code TALYS. The calculations reproduce forward angle scattering but have difficulty with relative minima in the differential cross section and large-angle scattering.     

Prompt gamma neutron activation analysis of boron with a 241Am-Be neutron source

A prompt gamma neutron activation analysis (PGAA) setup installed at ANRTC has been used to analyze boron. It consists of a 22.6% REGe detector and a 740 GBq ²⁴¹Am-Be neutron source moderated with water and paraffin. At the sample irradiation position, the thermal neutron fluence rate measured was 2.36·10⁴ n·m^–2· s^–1 and the corresponding Cd-ratio was 22 for gold monitor. The absolute detection efficiency in the range of 120–1500 keV was determined using ¹⁵²Eu standard solution. The sensitivity and detection limit for standard boric acid samples has been determined. The boron content in boric acid prepared from Turkish borate ores is measured to be 15.91±0.46% wt.     

Multi-geometry241AmBe neutron irradiator: Design and calibration for total-body neutron activation analysis

Neutron activation analysis provides an accurate elemental assessment of body compositional compartments; the analytical technique identifies skeletal, muscle, protein, and fat masses. Our irradiator system uses 56²⁴¹AmBe sources (4.4·10⁷ n/s each) which can be arranged in four basic geometries to examine body sizes from a preterm infant (500 g, 30 cm) to a very large adult (120 kg, 2m). Both in vivo and human cadaver studies have been performed. Precisions of 1 to 2% for total body Ca, P, Na, and Cl have been obtained.