Determination of thermal to fast neutron flux ratio (f(fast)) and fast neutron flux (ϕ(fast)) is required for fast neutron reactions, fast neutron activation analysis, and for correcting interference reactions. The f(fast) and subsequently ϕ(fast) were determined using the absolute method. The f(fast) ranged from 48 to 155, and the ϕ(fast) was found in the range 1.03×10(10)-4.89×10(10) n cm(-2) s(-1). These values indicate an acceptable conformity and applicable for installation of the fast neutron facility at the MNA research reactor.
A bulk of used paper supplied to recycling industry may contain water in their internal voids. This is because the price of the used paper is currently based on their weight and it has a huge potential of suppliers to add with water in order to increase the price. Currently used methods for detecting moisture content in a paper are restricted to a sheet of paper only. This paper presents a non-intrusive method for quick and in-situ measurement of water content in a bulk of used paper. The proposed method extends the capability of common paper moisture gauge, by using a neutron device. A fast neutron source (Am-Be 241) and a portable backscattering neutron detector are used for water measurement. It theoretically indicates that the slow neutron counts can be correlated to the hydrogen or water level in a paper. The method has the potential of being used by the paper-recycling industry for rapid and non-destructive measurement of water in a bulk of used paper.
Paper recycling plants usually buy their raw material from suppliers. More than often, bulk used paper supplied to the plant contains some significant quantity of water in its internal voids. It may be included intentionally or unintentionally. The price of used paper depends on its weight, thus adding water will help to increase weight and consequently increase the price. In this way, plant owner who purchase the used paper suffers a significant of financial lost. The objectives of our experiment are to establish a calibration curve that correlate between the amount of neutron backscattered and water content, and finally to develop a correction factor that need to be introduced to the measured values of water content. A fast neutron source (Am-Be 241) and a portable backscattering neutron detector were used for water measurement. The experiments were carried out by measuring neutron backscattering from used paper that has been added with different amount of water. As a result, a neutron calibration curve that provides a correlation between neutron backscattering and water content was established.
This work main aim is to study the analysis of slow neutrons which include thermal and
epithermal neutrons and also analysis on fast neutrons. The outcome from this work showed that
the comparison result between fast and slow neutrons. The safety assessment at reactor TRIGA
FUSFATI (RTF) is one of the main objectives of the work and there is a detailed discussion on it
which helped in accomplishing the task. Gamma Rays produced in this experiment was high and in
the experiment and it is realized that the shielding plays a vital role in the success of this
experiment which prevents all the radiations. From the results of the experiment it is realized that
these gamma rays are not suitable for the application of Boron Neutron Capture Therapy
(BNCT). However, these radiations are suitable for the application of Neutron Radiography (NR).
The study on this work will help in study of nuclear applications such as BNCT, NR, SANS etc.
These applications are using in medical and nuclear fields. The electronic device used in the
experiment to detect neutron is Neutron Spectrometer. The results from Neutron Spectrometer
and TLDs are very similar which showed that the experiment is a success. Numerical results were
compared with those available in literature for validation.
In industrial plants such as electricity generating, petroleum, chemical and petrochemical plants, pipelines are used extensively to transport liquid from one location to another. In radiation vulcanization of natural rubber latex (RVNRL) plants, pipelines are also used to transport latex to storage tank. During one of its maintenance activities, a pipeline intelligent gauge (PIG) that was used to monitor pipe integrity jammed inside the pipe causing interruption to its operation or loading activities. Sealed source technology was utilized to determine the location of jammed PIG in the pipeline. Fast neutrons from a 50 mCi Americium Beryllium (AmBe241), with energy range between 0.5 to11 MeV, were used for the study. Helium 3 (He3) detector was used to detect slow neutrons having a range of energy of 30 eV- 0.5 MeV. The investigation was carried out using neutron backscatter technique scanner. By adopting back-scattered technique, the location of jammed PIG in the pipeline has been successfully determined.
Nuclear radiation shielding capabilities for a glass series 20Bi2O3 - xPbO - (80 - 2x)B2O3 - xGeO2 (where x = 5, 10, 20, and 30 mol%) have been investigated using the Phy-X/PSD software and Monte Carlo N-Particle transport code. The mass attenuation coefficients (μm) of selected samples have been estimated through XCOM dependent Phy-X/PSD program and MCNP-5 code in the photon-energy range 0.015-15 MeV. So obtained μm values are used to calculate other γ-ray shielding parameters such as half-value layer (HVL), mean-free-path (MFP), etc. The calculated μm values were found to be 71.20 cm2/g, 76.03 cm2/g, 84.24 cm2/g, and 90.94 cm2/g for four glasses S1 to S4, respectively. The effective atomic number (Zeff)values vary between 69.87 and 17.11 for S1 or 75.66 and 29.11 for S4 over 0.05-15 MeV of photon-energy. Sample S4, which has a larger PbO/GeO2 of 30 mol% in the bismuth-borate glass, possesses the lowest MFP and HVL, providing higher radiation protection efficiency compared to all other combinations. It shows outperformance while compared the calculated parameters (HVL and MFP) with the commercial shielding glasses, different alloys, polymers, standard shielding concretes, and ceramics. Geometric Progression (G-P) was applied for evaluating the energy absorption and exposure buildup factors at energies 0.015-15 MeV with penetration depths up to 40 mfp. The buildup factors showed dependence on the MFP and photon-energy as well. The studied samples' neutron shielding behavior was also evaluated by calculating the fast neutron removal cross-section (ΣR), i.e. found to be 0.139 cm-1 for S1, 0.133 cm-1 for S2, 0.128 cm-1 for S3, and 0.12 cm-1 for S4. The results reveal a great potential for using a glass composite sample S4 in radiation protection applications.