The Reaction Of The BF3 Detector To Thermal Neutron
Three different reflecting materials, namely aluminum, wood, and Perspex of differing thicknesses, were used to assess the response of the BF3 detector to a 3 Ci Am–Be neutron source. Experiments have also been conducted to assess the varied contribution of the wall effect on the response owing to changes in the active volume of the detector. Then, utilizing source biasing and significance sampling, a Monte Carlo method was created to calculate the BF3 detector's neutron response function. In a three-dimensional source, detector, and reflecting medium arrangement, this code replicates the BF3 detector response exposed to the neutron field. The simulation findings were compared to the matching experimental data, and they were found to accord well. The experimental neutron albedo measurements for various Perspex thicknesses demonstrate saturating behavior, and the findings match the Monte Carlo simulation data quite well.
Neutron detectors have been extensively studied due to their use in the quantification of nuclear materials and shielding. The response of these detectors has been a source of concern, prompting a number of theoretical and experimental studies. For neutral and charged particle transport modeling, the Monte Carlo approach has been demonstrated to be one of the most accurate methodologies. Fixed sources and eigenvalue/criticality problems are the two types of nuclear problems to which the method is applied. Radiation transfer via a material can produce absorption and isotropic/anisotropic scattering in fixed source situations. Changes in energy and radiation direction result from these processes. The Monte Carlo approach has also been used to predict the response of neutron detectors theoretically. Experiments and traditional Monte Carlo simulations were used to investigate the features of proportional counters. Lippold et al. ran Monte Carlo simulations for neutron moisture gauges, Lees et al. (1980) for the BF3 detector assembly, and Dhairyawan et al. (1980) for spherically moderated neutron detectors, while Dhairyawan et al. (1980) for spherically moderated neutron detectors. In addition, a spherical neutron counter with a polyethylene moderator and three slim 3He position–sensitive proportional counters was designed for spectrometry and dosimetry. Dunn and Shultis (2009) have reviewed the use of traditional Monte Carlo methods for the design and analysis of radiation detectors in depth.
The energy responses of three types of Japanese neutron dose-equivalent meters were then derived using basic Monte Carlo calculations and experimental observations by Saegusa et al. (in 2004). Thermal neutrons, monoenergetic neutrons with energies up to 15.2 MeV, and neutrons from radionuclide sources like 252Cf and 241Am–Be have had their energy responses studied.
Hybrid Monte Carlo methods, which combine the fundamental Monte Carlo approach with empirical relations in the particle history within the detector, have recently emerged as rapid converging strategies for simulating stochastic response of detectors . When compared to traditional Monte Carlo processes, these approaches are more efficient and have faster convergence times.
The goal of this research is to look at how different reflector materials affect the BF3 detector's response function. First, the experimental response of the BF3 detector was investigated, as well as the relationship between the BF3 detector response and detector size and reflector thickness. Then, utilizing significance sampling in the usual Monte Carlo approach, a Monte Carlo–based computer code was built to simulate the response of the BF3 detector by considering the neutron histories in the BF3 detector. MATLAB version 6 was used to create this simulation application. MathWorks Inc. in the United States created this simulation program. The experimental setup and approach employed in this study are detailed in detail in section 2. The mathematical specifics of the model, including the governing differential equations, are then presented in section 3, along with details on computer implementations. In section 4, we describe how the findings produced using this technique compare to experimental data. In section 5, the study's findings are reported.
Materials and technique of experimentation
The BF3 detectors contain a small diameter center wire anode and a cylindrical outer cathode. Because of its low neutron interaction cross section, aluminum is employed as a cathode. At operational voltage, typical gas multiplication is on the order of 100–500. Because pulse height resolution degrades above room temperatures, these devices are restricted to temperatures up to around 100°C.
Measurements and setup
The neutron emission rate of 7.2 106 neutrons/second was achieved using a 3 Curie cylindrical Am–Be neutron source. At a depth of 11 cm from the top level, the source was encased in a cylinder filled with paraffin wax. Paraffin wax was utilized to thermalize neutrons released from the neutron source. The cylinder housing the source was covered with a slab about 7 cm thick. A passage on the paraffin surface was available for the positioning of the detector on the slab.
Procedure for simulation
A conventional cylindrical BF3 detector has been considered in this study. Table 1 shows the design characteristics of three distinct detectors that were employed. The BF3 detector response was calculated using neutron tracking from source to detector for a cylindrical Am–Be source. Kalos and Whitlock reported how they sampled the place of neutron emission, their direction of emission, and the free flight distance.
The flow chart depicting neutron transit from the source to the BF3 detector. The n(10B,7Li)4He reaction occurs when a neutron interacts with another neutron inside the detector.