Zohreh Gholamzadeh; Amir Pourrostam; Reza Ebrahimzadeh; Zeinab Naghshnejad
Abstract
In many human diseases and health cases, therapy of blood transfusion becomes necessary. In spite of the necessity, there are some risks associated with blood used in blood transfusion process. The TA-GVHD (transfusion-associated graft-versus-host-disease) is a problem when a blood transfusion occurs. ...
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In many human diseases and health cases, therapy of blood transfusion becomes necessary. In spite of the necessity, there are some risks associated with blood used in blood transfusion process. The TA-GVHD (transfusion-associated graft-versus-host-disease) is a problem when a blood transfusion occurs. The blood irradiation with gamma rays in blood bags can eliminate this risk. It should be mentioned that Co-60 sources are widely used for such blood irradiators. The present work investigates Co-60 production yield inside the external irradiation boxes of Tehran Research Reactor (TRR) using MCNPX code. 10-rod and 4-rod Co-59 assemblies were modeled at different external irradiation boxes to investigate their negative reactivity impact on TRR core as well Co-60 buildup rate during 3 years operation of the nuclear core at 4 MW power. The obtained results from MCNPX code showed a 4-rod assembly in linear form could obtain the highest specific activity (Ci.g-1) inside the external irradiation box faced to the core center. The computational results showed about 8 kCi of Co-60 is produced at the optimized irradiation position after 3 years TRR operation at 4 MW power.
Hadi Zanganeh; Mahdi Nasri Nasrabadi
Abstract
In this work, neutron and gamma shielding were simulated using MCNPX code for an inertial electrostatic confinement Fusion (IECF) device. In this regard, various properties of shields were investigated. Portland reinforced concrete was considered as the first layer. In addition to being effective in ...
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In this work, neutron and gamma shielding were simulated using MCNPX code for an inertial electrostatic confinement Fusion (IECF) device. In this regard, various properties of shields were investigated. Portland reinforced concrete was considered as the first layer. In addition to being effective in reducing the dosage of fast neutrons, concrete layer was also considerably effective in reducing the dose of gamma rays. As for the second and third layers, we opted for paraffin and boric acid based. These layers were chosen based on parameters such as lethargy, macroscopic slowing down power (MSDP), etc. in order to reduce the speed of epithermal neutrons and then absorb the thermal neutrons, thus reducing the transmitted neutron dosage as much as possible. A layer lead was used after these three layers of shielding to attenuate the gamma ray reaching this layer. In this study, a fusion source based on D-T fuel with homogeneous and isotropic radiation of neutrons was used and then dosimetry was performed for different parts. Afterwards, the thickness of the shielding layers was optimized in such a way that the neutron and gamma doses were reduced according to the standards. We found that it is possible to achieve safe neutron and gamma fluxes and doses by applying about 5 layers of 50 cm thickness. We compared the results of our study with the those of another study done on shielding for the IECF device, which were in good agreement.
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.
