Neutron spectroscopy with TENIS using an artificial neural network.

In this research, a ThErmal Neutron Imaging System (TENIS) consisting of two perpendicular sets of plastic scintillator arrays for boron neutron capture therapy (BNCT) application has been investigated in a completely different approach for neutron energy spectrum unfolding. TENIS provides a thermal neutron map based on the detection of 2.22 MeV gamma-rays resulting from 1H(nth, γ)2D reactions, but in the present study, the 70-pixel thermal neutron images have been used as input data for unfolding the energy spectrum of incident neutrons. Having generated the thermal neutron images for 109 incident mono-energetic neutrons, a 70 × 109 response matrix has been generated using the MCNPX2.6 code for feeding into the artificial neural network tools of MATLAB. The errors of the final results for mono-energetic neutron sources are less than 10% and the root mean square error (RMSE) for the unfolded neutron spectrum of 252Cf is about 0.01. The agreement of the unfolding results for mono-energetic and 252Cf neutron sources confirms the performance of the TENIS system as a neutron spectrometer.

Modeling GE advance PET-scanner using FLUKA simulation code.

The general-purpose particle transport code, FLUKA, was used to model the GE Advance Positron Emission Tomography (PET) scanner (GEMS). The required spatial and energy data were extracted using the FLUKA phase-space card, USERDUMP, with its user-routine, mgdraw, consisting of full-events information. Several point and volumetric sources were modeled and the corresponding locations of the sources were identified using three different proposed algorithms which eliminate the use of conventional image-reconstruction packages. Also, having extracted the time information, the time-of-flight approach in PET (TOF-PET) was used to reconstruct the images of several volumetric positron-emitting sources, representing the best results according to the chi-square analysis data. The post-processing calculations, both on CPU and GPU, were undertaken either with MATLAB or Fortran programming.

TENIS - ThErmal Neutron Imaging System for use in BNCT.

A thermal neutron flux measurement tool with two perpendicular sets of plastic scintillator arrays was designed and simulated (Ghal-Eh and Green, 2016) with the MCNPX code (Version 2.6.0, with ENDF/B-VII cross section library (ENDF, 2011)). The proposed system aimed to provide a thermal neutron map based on the detection of 2.22 MeV gamma-rays resulting from 1H(nth, γ)2D reactions. In the present work, using Monte Carlo code FLUKA and its scintillation light transport capability, several important upgrades were carried out to include the light transport modeling in the response of plastic scintillators, analyze the cross-talk phenomenon, optimize the system geometry, and also provide a new approach in thermal neutron image reconstruction. The results showed that the last two cases played a significant role in improving the longitudinal profile of thermal neutron flux.

Conceptual design for a new heterogeneous 241Am-9Be neutron source assembly using SOURCES4C-MCNPX hybrid simulations.

In this study, an approach to simulate a novel variable-yield heterogeneous 241Am-9Be was proposed with a hybrid use of SOURCES4C and MCNPX codes, and its energy spectrum and neutron emission yield were simulated. In these simulations, the energy spectra of the alpha particles emitted from the americium source and the neutrons produced within the beryllium and oxygen contents as a result of 9Be(α,n) and 17,18O(α,n) reactions were calculated with SOURCES4C whilst the neutron transport from neutron production points to the space outside the source assembly were performed with the MCNPX code. The neutron energy spectrum and neutron emission yield for two different configurations of single-rod and multi-rod sources (i.e., americium or americium oxide rods in beryllium medium) were compared to a source of homogeneous americium (or its oxides) and beryllium mixture. The proposed heterogeneous geometry was aimed to provide a neutron source with a variable neutron yield, easy-to-shut down and easy-to-waste process features. The results confirmed that the homogeneous source represented the largest neutron yield compared to single- and multi-rod geometries. However, the neutron yield in heterogenous geometry could be altered by varying the number of americium (or americium oxide) rods to reach the desired neutron yield.

A fuel for generation IV nuclear energy system: Isotopic composition and radiation characteristics.

This paper reports on an important issue of designing appropriate nuclear fuel of a high-temperature gas-cooled nuclear reactor operating in a thorium-plutonium nuclear fuel cycle. The neutronic calculations for a fuel of specific isotopic composition were performed before the analyses were done on the alpha emission probabilities, and on the neutron and photon sources as a result of (α,n) reaction. The main focus was on the quantitative evaluation of the neutron yield and the neutron energy spectrum for the generated neutrons through (α,n) reaction on light nuclei of dispersed nuclear fuel. Tests were carried out with the aim of creating an efficient calculation tool for the initial evaluation of the radiation characteristics for the irradiated multilayer nuclear fuel with different configurations and compositions.

A neutron scattering soil moisture measurement system with a linear response.

A prototype moisture measurement system was designed and constructed, based on neutron scattering, for preforming measurements in the laboratory. The system consisted of a rectangular soil container, an 241Am/Be neutron source and two parallel 10BF3 detectors (one near the source and the other far from it). Neutrons from the source are moderated and backscattered within the soil sample before being detected by two parallel counters, whose count ratios are shown to be linearly related to the soil moisture even within short measurement times. The system's performance was demonstrated using the Monte Carlo simulations, and a series of measurements on soil samples made of clay (40 wt%) and sand (60 wt%), mixed with different percentages of water. The results showed that the detectors response ratio is linear, up to about 25% of water content.

Replacement of Bonner spheres with polyethylene cylinders for the unfolding of an 241Am-Be neutron energy spectrum.

In this study, a conventional Bonner sphere spectrometer, together with 6LiI(Eu) inorganic scintillator used as the central detector, was used to obtain the neutron energy spectrum of an 241Am-Be source. To achieve this, we calculated the response matrices corresponding to eight different sizes of polyethylene spheres for the neutron energies ranging from about 10-7 to 15MeV in 54 energy groups by using the MCNPX2.6 code, and the 241Am-Be neutron spectrum was obtained using a modified version of the neutron spectrum unfolding code, AFITBUNKI. In a feasibility study, similar calculations were performed with different sizes of polyethylene cylinders. A comparison between the unfolded spectra of a typical 241Am-Be neutron source with those of Bonner spheres, cylinders of similar sizes, and standard 241Am-Be neutron energy spectra shows that the spectrometer with polyethylene cylinders can be used as a potential alternative for conventional Bonner spheres.

A plastic scintillator-based 2D thermal neutron mapping system for use in BNCT studies.

In this study, a scintillator-based measurement instrument is proposed which is capable of measuring a two-dimensional map of thermal neutrons within a phantom based on the detection of 2.22MeV gamma rays generated via nth+H→D+γ reaction. The proposed instrument locates around a small rectangular water phantom (14cm×15cm×20cm) used in Birmingham BNCT facility. The whole system has been simulated using MCNPX 2.6. The results confirm that the thermal flux peaks somewhere between 2cm and 4cm distance from the system entrance which is in agreement with previous studies.

Modelling boron-lined proportional counter response to neutrons.

The detailed Monte Carlo simulation of a boron-lined proportional counter response to a neutron source has been presented. The MCNP4C and experimental data on different source-moderator geometries have been given for comparison. The influence of different irradiation geometries and boron-lining thicknesses on the detector response has been studied.

A comparison study on three different radiation detectors used for liquid levelmetry.

In this paper, three different radiation detectors (BF3 counter, NE213 and BGO scintillators) and an (241)Am-Be isotopic neutron-gamma source have been used for a typical liquid levelmetry. The study shows that the use of the Am-Be source together with an NE213 scintillator has the best performance.

Neutron-gamma discrimination with UGAB scintillator using zero-crossing method.

The new-type scintillator, Ultima Gold Alpha-Beta (UGAB), was studied for its neutron-gamma discrimination capability. The figure-of-merit and peak-to-valley values for the neutron-gamma discrimination spectra of UGAB scintillator when exposed to (241)Am-Be neutron source were presented. The results show that this new-type scintillator can efficiently be used in neutron-gamma discrimination experiments.

A simple well-logging tool using boron-lined sodium iodide scintillators and an 241Am-Be neutron source.

A couple of 2-inch by 2-inch right cylinder sodium iodide scintillators and an Am-Be radioisotope neutron source have been used in a neutron porosity well-logging tool to explore the variation of hydrogen contents in a prototype formation. Both Monte Carlo N-particle transport code simulation and experimental results of the near- to far-detector responses confirm the reliable sensitivity of proposed tool to the formation porosity.

Modelling plastic scintillator response to gamma rays using light transport incorporated FLUKA code.

The response function of NE102 plastic scintillator to gamma rays has been simulated using a joint FLUKA+PHOTRACK Monte Carlo code. The multi-purpose particle transport code, FLUKA, has been responsible for gamma transport whilst the light transport code, PHOTRACK, has simulated the transport of scintillation photons through scintillator and lightguide. The simulation results of plastic scintillator with/without light guides of different surface coverings have been successfully verified with experiments.

Photocathode non-uniformity contribution to the energy resolution of scintillators.

This paper introduces the basics of the light transport simulation in scintillators and the wavelength-dependencies in the process. The non-uniformity measurement of the photocathode surface is undertaken, showing that for the photocathode used in this study the quantum efficiency falls to about 4 % of its maximum value, especially in areas far from the centre. The wavelength- and position-dependent quantum efficiency is implemented in the Monte Carlo light transport code, showing that, the contribution of the photocathode non-uniformity to the energy resolution is estimated to be around 18 %, when all position- and wavelength-dependencies are included.

Light transport feature for SCINFUL.

An extended version of the scintillator response function prediction code SCINFUL has been developed by incorporating PHOTRACK, a Monte Carlo light transport code. Comparisons of calculated and experimental results for organic scintillators exposed to neutrons show that the extended code improves the predictive capability of SCINFUL.