Evaluation of neutron and gamma dose in a new deuterium-deuterium fusion neutron generator facility using MCNP and experimental methods.

A deuterium-deuterium (D-D) neutron generator facility has been developed with a wide range of applications in mind. D-D neutron generators provide a safe, convenient, and low-cost solution for the production of thermal, epithermal, and fast neutrons for laboratory and industrial applications. The regulatory burden for operating one of these generators is small as it contains no radioactive material and produces no radiation when turned off. One such D-D neutron generator has been installed at the National Center of Nuclear Sciences and Technologies (CNSTN). This generator produces 2.45 MeV fast neutrons with a maximum yield of 1010.n/s. To maintain a strong radiation safety program, detailed knowledge of the radiation dose rates around the neutron generator is crucial for ensuring the radiological protection of the personnel involved with its operation. This work describes the Monte Carlo calculations and experimental measurements carried out in the Neutron Generator facility of the CNSTN. The adequacy of the radiation shielding of the facility was verified by measuring the neutron and gamma dose rates at various locations inside and outside the neutron generator hall during operation and comparing them with Monte Carlo calculations. Measured and calculated dose rates were in agreement to better than 20% depending on location. The operation of this neutron facility will provide a suitable neutron source for future basic and applied research.

Investigation of flow dynamics of phosphate fertilizer production reactors using radiotracer technique.

Radiotracer investigations were carried out in two identical phosphoric acid production reactors and a triple super phosphate (TSP) production reactor in three different plants in Tunisia. The main objective of the study was to investigate and compare their flow behavior and identify flow abnormalities, if any. Residence time distribution (RTD) of the process material (ore pulp) was measured in the three reactors using Iodine-131. The measured RTDs were treated and analyzed to obtain flow parameters such as the mean residence time (MRT), dead volume, and extent of bypassing. The treated RTD curves were modeled using a suitable mathematical model, and the values of the parameters were obtained. The results of the investigations were used to evaluate and compare the flow performance of the reactors, quantify the degree of mixing, and visualize the prevailing flow patterns. The results of the investigations are expected to be used to make necessary modifications to enhance the efficiency and optimize the performance of the reactors or the production process.

Efficiency transfer using the GEANT4 code of CERN for HPGe gamma spectrometry.

In this work we apply the GEANT4 code of CERN to calculate the peak efficiency in High Pure Germanium (HPGe) gamma spectrometry using three different procedures. The first is a direct calculation. The second corresponds to the usual case of efficiency transfer between two different configurations at constant emission energy assuming a reference point detection configuration and the third, a new procedure, consists on the transfer of the peak efficiency between two detection configurations emitting the gamma ray in different energies assuming a "virtual" reference point detection configuration. No pre-optimization of the detector geometrical characteristics was performed before the transfer to test the ability of the efficiency transfer to reduce the effect of the ignorance on their real magnitude on the quality of the transferred efficiency. The obtained and measured efficiencies were found in good agreement for the two investigated methods of efficiency transfer. The obtained agreement proves that Monte Carlo method and especially the GEANT4 code constitute an efficient tool to obtain accurate detection efficiency values. The second investigated efficiency transfer procedure is useful to calibrate the HPGe gamma detector for any emission energy value for a voluminous source using one point source detection efficiency emitting in a different energy as a reference efficiency. The calculations preformed in this work were applied to the measurement exercise of the EUROMET428 project. A measurement exercise where an evaluation of the full energy peak efficiencies in the energy range 60-2000 keV for a typical coaxial p-type HpGe detector and several types of source configuration: point sources located at various distances from the detector and a cylindrical box containing three matrices was performed.

Development of a new deuterium-deuterium (D-D) neutron generator for prompt gamma-ray neutron activation analysis.

A new deuterium-deuterium (D-D) neutron generator has been developed by Adelphi Technology for prompt gamma neutron activation analysis (PGNAA), neutron activation analysis (NAA), and fast neutron radiography. The generator makes an excellent fast, intermediate, and thermal neutron source for laboratories and industrial applications that require the safe production of neutrons, a small footprint, low cost, and small regulatory burden. The generator has three major components: a Radio Frequency Induction Ion Source, a Secondary Electron Shroud, and a Diode Accelerator Structure and Target. Monoenergetic neutrons (2.5MeV) are produced with a yield of 10(10)n/s using 25-50mA of deuterium ion beam current and 125kV of acceleration voltage. The present study characterizes the performance of the neutron generator with respect to neutron yield, neutron production efficiency, and the ionic current as a function of the acceleration voltage at various RF powers. In addition the Monte Carlo N-Particle Transport (MCNP) simulation code was used to optimize the setup with respect to thermal flux and radiation protection.

Monte Carlo simulation of explosive detection system based on a Deuterium-Deuterium (D-D) neutron generator.

An explosive detection system based on a Deuterium-Deuterium (D-D) neutron generator has been simulated using the Monte Carlo N-Particle Transport Code (MCNP5). Nuclear-based explosive detection methods can detect explosives by identifying their elemental components, especially nitrogen. Thermal neutron capture reactions have been used for detecting prompt gamma emission (10.82MeV) following radiative neutron capture by (14)N nuclei. The explosive detection system was built based on a fully high-voltage-shielded, axial D-D neutron generator with a radio frequency (RF) driven ion source and nominal yield of about 10(10) fast neutrons per second (E=2.5MeV). Polyethylene and paraffin were used as moderators with borated polyethylene and lead as neutron and gamma ray shielding, respectively. The shape and the thickness of the moderators and shields are optimized to produce the highest thermal neutron flux at the position of the explosive and the minimum total dose at the outer surfaces of the explosive detection system walls. In addition, simulation of the response functions of NaI, BGO, and LaBr3-based γ-ray detectors to different explosives is described.

Characterization of three digital signal processor systems used in gamma ray spectrometry.

Various manufacturers have recently introduced digital signal processing systems that allow data acquisition in gamma spectrometry at high-input counting rates (several thousand pulses per second). In these systems, the signal digitization is performed immediately following the preamplification stage. This allows digital shaping and filtering of the signal which increases the number of possible combinations in signal shaping and as a consequence, optimizes the resolution as a function of the detector characteristics and the counting rate. Basic characteristic parameters of three digital signal processors that were recently introduced in the market have been studied and compared to those of an analog system. This study is carried out using a hyper-pure coaxial type germanium detector and 57Co, 60Co and 137Cs radioactive sources. Performance parameters such as energy resolution, system throughput, and counting losses that are due to dead time and pile-up effects are presented and discussed.