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.
S. Taher Aminfarkhani; Ahmad Lashkari; S. Farhad Masoudi
Abstract
The present work is concerned on neutron flux increasing in Tehran Research Reactor (TRR). TRR is a 5 MW pool-type research reactor with plate type fuels in which the light water is used as both the coolant and moderator. The main goal of this paper is reaching to the average thermal neutron flux of ...
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The present work is concerned on neutron flux increasing in Tehran Research Reactor (TRR). TRR is a 5 MW pool-type research reactor with plate type fuels in which the light water is used as both the coolant and moderator. The main goal of this paper is reaching to the average thermal neutron flux of the order of 1014 #/cm-2.s-1 in the central irradiation box. Combination of the TRR power upgrading with the compact core can enable us to reach a neutron flux higher than 1.5 × 1014 #/cm-2.s-1 without violating the neutronic and thermal-hydraulic safety criteria. The compact core, with 19 and 5 standard and control fuel elements respectively, is used as a base for the neutronic analyses. Compact core with 26 fuel assemblies fulfilled all neutronic and operation criteria. Considering thermal hydraulic aspect from previous study lets TRR to be upgraded to 8.5 MW, resulting in neutron thermal flux greater than 1.5 × 1014
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%.
Afshin Hedayat
Abstract
Fast neutron irradiation is one of the most strategic radiation applications of research reactors. Usually, it is performed around the reactor core containing lower neutron flux. In this paper, a hybrid object has been introduced and analyzed to enhance irradiating ...
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Fast neutron irradiation is one of the most strategic radiation applications of research reactors. Usually, it is performed around the reactor core containing lower neutron flux. In this paper, a hybrid object has been introduced and analyzed to enhance irradiating applications of the fast neutrons in the core of a Material Testing Reactor (MTR). The tool includes an old-type low-consumed HEU control fuel element, a dry channel, and a Cd filter. It is supposed to be installed at the internal neutron trap (D4 positions) of TRR core configuration. Calculating results are very promising for using the proposed tool to increase neutron fluxes, reduce thermal and epi-thermal neutron fluxes, and shift the neutron spectrum toward the fast neutron region (hardening effect) at the chosen irradiating location. Primary safety parameters are also checked and passed successfully. Furthermore, there are also some other presented safety items which must be checked carefully and conservatively in order to refabricate and install such a irradiating tool in an in-core location of a MTR.
