Zohreh Gholamzadeh; Reza Ebrahimzadeh; Mohammad Hossein Choopan Dastjerdi; Javad Mokhtari
Abstract
New nitrile butadiene rubber (NBR) materials are being considered to use for neutron shielding especially for the positions which needs a flexible neutron shield. Such light, low-cost, and suitable material could be used for sealing of the gaps or even for shielding of low radiation environments. In ...
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New nitrile butadiene rubber (NBR) materials are being considered to use for neutron shielding especially for the positions which needs a flexible neutron shield. Such light, low-cost, and suitable material could be used for sealing of the gaps or even for shielding of low radiation environments. In the present work, experimental investigation of NBR shielding performance of neutrons and gamma rays was proposed using the beam line of the Isfahan Miniature Neutron Source Reactor . MCNPX code was used to simulate the 30 kW research reactor beam line. Six NBR sheet with 2 cm thickness were used at the outlet of the beam line respectively to measure its neutron shielding as well as gamma shielding power on thickness. The experiment situations were modeled using the computational code. The obtained results showed the flexible and cheap material could be used as a good neutron shield while it acts as a very weak shield for gamma rays too. Also there is good conformity between simulation and experimental data with maximum 37% relative discrepancy.
Zohreh Gholamzadeh; Rohollah Adeli; Mahdi Keivani
Abstract
Routine gamma dosimetry of spent fuels in nuclear power stations is mandatory to manage their storage in dry or wet spent fuel storages. Mostly the spent fuel gamma dose rate measurements out of the spent fuel pool is impossible because of the high exposures of the operators. Therefore, ...
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Routine gamma dosimetry of spent fuels in nuclear power stations is mandatory to manage their storage in dry or wet spent fuel storages. Mostly the spent fuel gamma dose rate measurements out of the spent fuel pool is impossible because of the high exposures of the operators. Therefore, determination of a conversion factor as precise as possible is important that could be applied to convert the measured gamma dose rate inside the water shield to the air values. Simulation methods are powerfully applied to investigate the conversion factor variation trends due to different burnup, cooling time and irradiation history of the spent fuels. The present work uses MCNPX Monte Carlo-based code to determine the trend. The obtained results of this computational study showed that the conversion factor would not have any dependency to the cooling times, burnup values and irradiation history if the detector is placed at special positions in air or water environments. Comparison of the simulation and experimental data showed an acceptable conformity, so that the experimental verified the simulation data trend
Zohreh Gholamzadeh; Mohadeseh Gholshanian; Seyed Mohammad Mirvakili
Abstract
Today thorium based fuels are being investigated as an alternative fuel technology. However, the majority of thorium fuel research studies are limited to reactor physics investigations, which leaves a gap for dose evaluation and shielding concerns of such spent fuels. The present work investigates thorium ...
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Today thorium based fuels are being investigated as an alternative fuel technology. However, the majority of thorium fuel research studies are limited to reactor physics investigations, which leaves a gap for dose evaluation and shielding concerns of such spent fuels. The present work investigates thorium oxide fuel assemblies in Tehran research reactor. The fuel gamma dose rates are calculated at different burnups and cooling times. A comparison between the reactor routine fuel and the thorium oxide fuel is conducted to reveal the thorium-based fuel application shielding challenges. The obtained results showed that inverse to U3O8-Al routine fuel the spent ThO2 gamma dose rates are completely dependent to the burnup values. In addition, for transporting the spent ThO2 fuel with the routine transport casks there is needed to be waited for the higher cooling times than U3O8-Al transportation time or construction of thicker transport casks is needed for transportation of the thorium-based spent fuels at shorter times.
