Majid Zamani; Mohsen Shayesteh
Abstract
Using the experimental data in nuclear computing to verify the calculation methods and tools based on numerical and statistical methods has many benefits such as illustrating the quality, ensuring the capabilities, and computer codes validating. Simulation by computer tools is also applicable in the ...
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Using the experimental data in nuclear computing to verify the calculation methods and tools based on numerical and statistical methods has many benefits such as illustrating the quality, ensuring the capabilities, and computer codes validating. Simulation by computer tools is also applicable in the safety analysis of research reactors. In this research, the computer tool (MCNPX 2.7.0: 2011) was verified against the experimental data of neutron flux and spectrum on the sample position of the Tehran Research Reactor (TRR) neutron imaging system by the neutron activation method. To determine the benchmark specifications, the simulation of the system was done at the first step by considering a well-defined facility geometric, material specification and reactor core configuration, fuel elements, and radiation facility (beam tubes and collimator, reactor core, and neutron imaging components). Then the flux and neutron spectrum at the sample position were calculated. In the second step, a set of In (bare and covered by cd) and Au foils and a set of Au, Ni, Ti, and Zr, were placed and exposed almost in front of the reactor E beam tube. The neutron energy spectrum was unfolded by calculating the saturation activity of each foil by SAND-II code, and the neutron flux was calculated. A comparison of the results obtained in two steps shows a relatively good and acceptable agreement (Max. 30% deviation) between the flux and the shape of the flux profile obtained from calculations and experimental data.
Mahya Pazoki; Hamid Jafari; Zohreh Gholamzadeh
Abstract
Neutron data and cross-sections are highly regarded and are essential for developing nuclear equipment such as advanced fission and fusion reactors, accelerators, neutron shielding, physics studies, etc. The neutron cross-section should preferably be measured using a single-energy neutron beam, although ...
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Neutron data and cross-sections are highly regarded and are essential for developing nuclear equipment such as advanced fission and fusion reactors, accelerators, neutron shielding, physics studies, etc. The neutron cross-section should preferably be measured using a single-energy neutron beam, although the presence of a background in research reactors can affect its accurate determination. The Neutron Powder Diffraction (NPD) facility of Tehran Research Reactor (TRR) has been taken into consideration for measuring the neutron cross-section based on its properties, including neutron monochromator and multiple collimators. In this work, radiative capture cross-sections of Au, In, and Rh materials have been calculated using TRR monochromatic beam. MCNPX is a Monte Carlo particle transport code that has been applied to simulate the measurement system of the neutron cross-section and calculate the reaction rates. The effect of the presence and absence of different sections of the background on the cross-section values was investigated and the results were compared with EXFOR data library for validation. According to the findings, neutron backgrounds can have varying impacts depending on factors such as sample material, the isotope resonance regions, neutron source spatial distribution, and neutron monochromatic energy. However, the presence of fast neutron background contributes to the most uncertainty in the cross section values while its removal produces an average discrepancy from experimental libraries of 7.16%. Also, removing the cold neutron background also causes a relative difference equal to 7.65%.
