Reza Pourimani; Mohammad Reza Zare; Mehrdad Aghamohammadi
Abstract
In this work, the concentration of tritium in D2O of various degrees of purity was measured. Samples were taken from the Arak heavy water plant and tritium concentrations were determined using a liquid scintillation detector (LSC) based on tritium decay. In this work, instead of simple distillation, ...
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In this work, the concentration of tritium in D2O of various degrees of purity was measured. Samples were taken from the Arak heavy water plant and tritium concentrations were determined using a liquid scintillation detector (LSC) based on tritium decay. In this work, instead of simple distillation, is used the azeotropic distillation method. Absorption and fluorescence spectra were recorded using a Shimadzu UV-2100 spectrometer and an LS50B fluorescence spectrometer. The tritium concentration in the samples varied from 1.75 ± 0.80 to 6.16 ± 1.01 Bq.L-1 in D2O enrichment from 0.35% to 77.50%. The correlation coefficient between tritium concentration and D2O purity in heavy water was obtained as R2 = 0.853. Deviation for 99.8% D2O enriched in heavy water. This was observed from a straight line, leading to a drop in R2. The results of this measurement showed that the tritium concentration did not exceed the value set by the Nuclear Regulatory Commission (NRC).
Mahdi Aghayan; S. Farhad Masoudi; Farshad Ghgasemi; Hamed Shaker
Abstract
Advantage of non-brazing methods in manufacturing of cavities has been considered in high gradient studies because of the softening of copper by brazing cavities at high temperatures. Recent studies with hard copper cavities have been shown that the harder materials can reach larger accelerating gradients ...
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Advantage of non-brazing methods in manufacturing of cavities has been considered in high gradient studies because of the softening of copper by brazing cavities at high temperatures. Recent studies with hard copper cavities have been shown that the harder materials can reach larger accelerating gradients for the same breakdown rate. Shrinking, as a braze-free method for construction of the cavities, was used recently to fabricate and assemble acceleration cavities of an electron linear accelerator at the Institute for Research in Fundamental Science (IPM-Iran). Based on the results obtained in this project, this paper proposes the design of a 3-cell S-band standing wave structure operating at 2.9985 GHz for high gradient tests, considering shrink-fit construction method. The desired cavity consists of three cells so that the maximum gradient in the middle cell is about twice that of the surrounding cells. Simulation with Ansys-HFSS showed that maximum axial electric field 59 MV.m-1 achievable for 2 MW input power in middle cell.
Mohammad Askari; Nikoo Darestani Farahani; Mehdi Bakhshzad Mahmoudi; Fereydoun Abbasi Davani
Abstract
Metal surface cleaning or etching techniques using reactive plasma are emerging as one of the dry processing techniques for surface contaminants with high bond energy, especially for cleaning and decontamination of nuclear components and equipment. In this study, the plasma reaction due to the discharge ...
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Metal surface cleaning or etching techniques using reactive plasma are emerging as one of the dry processing techniques for surface contaminants with high bond energy, especially for cleaning and decontamination of nuclear components and equipment. In this study, the plasma reaction due to the discharge of a dielectric barrier of a mixture of 95% helium and 5% fluorine with cobalt oxide film (Co3O4) grown on the surface of stainless steel 304 was studied experimentally. Experimental results show that cobalt oxide becomes a powder after plasma irradiation and is easily separated from the surface of the base metal. The optimal plasma generating conditions of the dielectric barrier discharge used in this experimental study were obtained at atmospheric pressure, voltage 4.5 kV, and frequency 25 kHz with an etching rate of 10.875 μmol.min-1. The samples were analyzed before and after plasma irradiation, using Scanning Electron Microscopy with Energy Dispersive X-ray spectroscopy and the purification rate was performed using a sequential weighting of the samples with scales 10-4 g accurately obtained. The results show the ability of this method to effectively remove the surface contamination of cobalt from the surface of stainless steel 304.
Payman Rafiepour; Shahab Sheibani; Daryiush Rezaey Uchbelagh; Hossein Poorbaygi
Abstract
Radioactive stents loaded with I-125 seeds have been widely used for the treatment of advanced esophageal cancer. Understanding the dose distribution of such stents before the clinical use is essential. This study provides a dosimetric investigation of I-125 seed-loaded stents based on the seed's arrangement ...
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Radioactive stents loaded with I-125 seeds have been widely used for the treatment of advanced esophageal cancer. Understanding the dose distribution of such stents before the clinical use is essential. This study provides a dosimetric investigation of I-125 seed-loaded stents based on the seed's arrangement and activity. A cylindrical water equivalent phantom with an esophageal stent loaded with I-125 seeds, were employed. The seeds arrangements were determined based on the distance between the centers of two adjacent seeds (z) along the stent length. EBT3 films as well as Geant4 Monte Carlo toolkit were used to obtain the dose distribution around the stent. By modeling the MIRD phantom, the dose delivered to the related organs at risk was calculated. The appropriate dose distribution is achieved for z=15 mm, in which the absorbed dose at a depth of 5 mm reaches about 45% of the absorbed dose near the stent surface, thereby the therapeutic dose is delivered to the reference points. Both arrangements (z=15 and 20 mm) seemed to be clinically eligible and their utilization depends on the patient and the hospital facilities. Using esophageal stents with z>20 mm is not recommended due to the presence of cold spots in the dose distribution.
Amir Moslehi; Mohammad Hossein Choopan Dastjerdi
Abstract
In the present work performance of film badge as an alternative personal dosimeter for thermal neutrons is investigated. To do this, a cadmium-lead (Cd-Pb) filter with the same thickness as the tin-lead (Sn-Pb) filter is attached to the AERE/RPS badge. Since thermal neutrons are mixed with gamma rays, ...
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In the present work performance of film badge as an alternative personal dosimeter for thermal neutrons is investigated. To do this, a cadmium-lead (Cd-Pb) filter with the same thickness as the tin-lead (Sn-Pb) filter is attached to the AERE/RPS badge. Since thermal neutrons are mixed with gamma rays, the dosimeter is irradiated by the 60Co gamma rays standard field of Karaj Secondary Standard Dosimetry Lab as well as the mixed neutron-gamma field of the radiography beamline of Isfahan Miniature Neutron Source Reactor. In the both fields, ten personal dose-equivalent values between 0.1 to 10 mSv are chosen. For any dose, three film badges are used and the net optical density is determined as the average of their optical densities. Finally, the calibration curves of the film badge are plotted to determine the dose-equivalent values. Obtained results reveal that film badge simultaneously determines the thermal neutrons and gamma rays dose fractions. Also, the thermal neutron doses are at most 50% different from the nominal values considered.
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
S. Farhad Masoudi; Fatemeh Sadat Rasouli
Abstract
Due to the selectively treating tumors and largely sparing normal neighboring cells, Boron Neutron Capture Therapy (BNCT) continues to be of special significance and interest for wide groups of researchers. One of the most important challenges in this context is to design an optimized ...
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Due to the selectively treating tumors and largely sparing normal neighboring cells, Boron Neutron Capture Therapy (BNCT) continues to be of special significance and interest for wide groups of researchers. One of the most important challenges in this context is to design an optimized beam based on an appropriate neutron source. The recent studies, focused on investigating neutron sources as alternatives for nuclear reactors, revealed the high potential of electron linac-based photoneutron sources to improve the efficiency of this treatment method. Inquiring about the efficiency of a layered model of beam shaping assembly (BSA) for photoneutron sources to be used in BNCT of deep tumors is the main subject of this simulation study. This model, unlike the traditional BSA in which the reflector surrounds the whole moderator, includes many concentric cylinders of reflectors and moderators. The MCNPX simulations for various primary energies show that the layered model results in more appropriate beam characteristics compared with that of the common geometry. Moreover, the parameters governing the beam properties such as the thickness of the layers, moderator/reflector and collimator lengths, and the thickness of the surrounding reflector have been investigated. The results are encouraging and offer new ways to accomplish more researches in studies on the BNCT technique.
Eugene Echeweozo; A.D Asiegbu; E.L. Efurumibe; L.A. Nnanna; H.K. Idu
Abstract
Gamma radiation shielding of baked and unbaked granite bricks produced with 0%, 10%, 20%, 30%, 40%, and 50% of kaolin powder were experimentally and theoretically assessed for possible deployment in liquid radioactive waste storage. A 3×3 ...
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Gamma radiation shielding of baked and unbaked granite bricks produced with 0%, 10%, 20%, 30%, 40%, and 50% of kaolin powder were experimentally and theoretically assessed for possible deployment in liquid radioactive waste storage. A 3×3 inches NaI(Ti) detector and WinXCOM program were used to measure the linear attenuation coefficients at different energies. Elements composition of samples were analyzed using particle induced X-ray emission (PIXE) spectroscopy. Results show that adding kaolin to granite positively reduced the liquid permeability coefficients of the bricks but negatively reduced the shielding properties of the bricks. Optimum results were obtained from unbaked sample of granite brick produced with 50% of micro scale kaolin powder (GK50) with mass attenuation coefficient of 0.0663, 0.0572 and 0.0552 cm2.g-1, radiation protection efficiency (RPE) of 38.36%, 34.11% and 33.13% for radiation energies levels of 661.6, 1173.2, and 1332.5 keV respectively and liquid permeability coefficient of 6.53×10-11 m.s-1. The study concludes that all brick samples were thermally stable, good in gamma radiation shielding and efficient in liquid radioactive waste immobilization.
Mohammad Ali Hejazi; Seyed Khalil Mousavian; Mohammad Outokesh
Abstract
In this study, thermal-hydraulic analysis of a dry storage cask for Bushehr Nuclear Power Plant spent nuclear fuels is carried out. Geometry drawing and mesh generation were completed in SolidWorks and Gambit software, respectively. Three different cases were considered for the cask geometry and design ...
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In this study, thermal-hydraulic analysis of a dry storage cask for Bushehr Nuclear Power Plant spent nuclear fuels is carried out. Geometry drawing and mesh generation were completed in SolidWorks and Gambit software, respectively. Three different cases were considered for the cask geometry and design including cask with/without spacers and cask with spacers and fins. Thermal-hydraulic analysis of the cask was performed for steady-state and normal storage conditions in ANSYS CFX solver package. Simulation results indicated a weak thermal-hydraulic behavior of the cask in the geometry without spacer and maximum fuel temperature exceeded the allowable safety limits. However, with the addition of spacers and fins in the geometry of the cask, thermal behavior of the cask was significantly improved and maximum fuel temperature achieved a proper margin compared to the allowable safety limits. As a result, the spent fuel integrity will be maintained in the normal storage conditions. The simulation results were compared with a literature published paper and it showed a good agreement between the calculated results.
Mohammadhosein Farzin; Mahdi Radin; Mahdi Moeini Arani
Abstract
We study the low-energy deuteron-deuteron elastic scattering using the cluster effective field theory formalism up to next-to-leading order (NLO). For this purpose, we initially focus on determination of the unknown effective field theory coupling constant values using the phase shift analysis and available ...
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We study the low-energy deuteron-deuteron elastic scattering using the cluster effective field theory formalism up to next-to-leading order (NLO). For this purpose, we initially focus on determination of the unknown effective field theory coupling constant values using the phase shift analysis and available differential cross section data. The differential cross section versus center of mass energy and scattering angle are plotted up to NLO in the suggested power counting and compared to the available experimental data. Our effective field theory results showgood consistency with the present data.
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.
Mohammad Hossein Bahrevar; Gholamreza Jahanfarnia; Ali Pazirandeh; Mohsen Shayesteh
Abstract
In this study, thermal-hydraulic analysis of partial loss of coolant flow accident in supercritical pressure light water reactor (SCWR) with a new geometric design has been investigated. In the new design, the coolant and moderator circuits are separated. This analysis was performed using the development ...
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In this study, thermal-hydraulic analysis of partial loss of coolant flow accident in supercritical pressure light water reactor (SCWR) with a new geometric design has been investigated. In the new design, the coolant and moderator circuits are separated. This analysis was performed using the development of a transient-state thermal-hydraulic code in which the equations of mass, momentum, and energy are solved. The porous Media approach is used to solve these equations. By extracting the results of transition modeling, it is observed that in the new geometric design, by separating the coolant and moderator circuits, the maximum fuel clad temperature is lower than the maximum fuel clad temperature value of the previous designs. As in the new design at the end of the transition, the maximum fuel clad temperature has decreased by about 37% compared to the initial state. The result of the calculations in this study shows that the new design, in which the coolant and moderator circuits are separated, has created more safety in a chosen transition.
Hamed Khodadadi; Amir Zareidoost
Abstract
After some accidents like TMI and Fukushima Daiichi, the belief that nuclear fuels can safely and reliably be used up to 65 MW.d/kgU was revised. Different short to long term concepts started to be assessed more seriously. Among these concepts, Cr-coated Zr cladding tubes were taken into consideration ...
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After some accidents like TMI and Fukushima Daiichi, the belief that nuclear fuels can safely and reliably be used up to 65 MW.d/kgU was revised. Different short to long term concepts started to be assessed more seriously. Among these concepts, Cr-coated Zr cladding tubes were taken into consideration as a promising remedy for the short term perspectives. Various efforts to evaluate important aspects of such designs (including coating methodologies, full size coating of cladding tubes, performance analysis, licensing features and so on) have been implemented and some activities are yet under accomplishment. In this study, an optimized cathodic arc PVD coating condition, from the defect density point of view, for applying Cr and CrN coating layers up 10 microns on typical VVER-1000 cladding tubes (Zr-1%Nb alloy) have been obtained. SEM results approved improvements, and furthur modification has been postponed till the finalizing the performance examinations in normal operational and accidental conditions (specifically, LB LOCA condition). Additionally, the list of required test, from characterizations to mechanical and performance experiments, in order to qualify the applicability of such coated samples are suggested.
Amir Charkhi; Parisa Zaheri; Amjad Sazgar; Iman Dehghan
Abstract
Since the production of tellurium hexafluoride gas requires the design of a suitable reactor system, so the study of tellurium oxide fluorination kinetics is of great importance. For this purpose, a novel laboratory system was designed and constructed to study the fluorination reactions by the volumetric ...
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Since the production of tellurium hexafluoride gas requires the design of a suitable reactor system, so the study of tellurium oxide fluorination kinetics is of great importance. For this purpose, a novel laboratory system was designed and constructed to study the fluorination reactions by the volumetric method. Fluorine gas was injected into the reactor containing a tellurium oxide pellet, and the reaction was studied by following the changes in pressure of the gas phase using a pressure transmitter instrument. In this volumetric system, the kinetic parameters of the reaction between tellurium oxide pellet and fluorine gas have been derived for a pressure range of 137.9 and 181.2 kPa by monitoring the gas phase pressure. The reaction temperature was adjusted to 204±1 ◦C using a heater. The results showed that the fluorination reaction of tellurium oxide is a first-order reaction. The reaction rate constant is calculated to be 6.86 × 10-4 s-1.
Fatemeh Sadat Rasouli
Abstract
Among the approaches commonly used to extend the sharp peak of the deposited dose in proton therapy, passive scattering is widely used and also is of concern because of the potential for generating secondary particles, especially neutrons, which can damage the non-target healthy tissues. The present ...
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Among the approaches commonly used to extend the sharp peak of the deposited dose in proton therapy, passive scattering is widely used and also is of concern because of the potential for generating secondary particles, especially neutrons, which can damage the non-target healthy tissues. The present simulation-based study investigates the effect of using the passive method for different primary proton energies on the dose delivered to the tissue compared with those of the pencil beam scanning method. The results show that the generation of secondary neutrons strongly depends on the material used in the beam design. Also, it was found that the passive method would lead to the physical neutron dose higher than that of the beam scanning method for various primary proton energies.
Ali Azizi Ganjgah; Payvand Taherparvar
Abstract
Radiation therapy aims to maximize doses to cancer cells while minimizing damage to normal tissues. Today, nanoparticles containing high-atomic-number elements, such as gold, gadolinium, and silver, have proven effective as radiosensitizers in radiotherapy to enhance dose delivery for cancer treatment. ...
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Radiation therapy aims to maximize doses to cancer cells while minimizing damage to normal tissues. Today, nanoparticles containing high-atomic-number elements, such as gold, gadolinium, and silver, have proven effective as radiosensitizers in radiotherapy to enhance dose delivery for cancer treatment. In this study, we used the Geant4-DNA toolkit to investigate the effects of multiple nanoparticles (NPs) with varying sizes (radius= 3.15 to 5 nm) on DNA damage when exposed to monoenergetic photons with energies of 15, 40, 50, and 68 keV. Direct and indirect single-strand breaks (SSBs), double-strand breaks (DSBs), and hybrid double-strand breaks (Hybrid DSBs) were calculated in the presence and absence of 1 to 4 nanoparticles (NPs) of the same total volume of gold, gadolinium, and silver nanoparticles for the 1ZBB model (selected from the Protein Data Bank (PDB) library). The results show that increasing the number of gold, gadolinium, and silver NPs and decreasing the photon beam energy increases the total number of strand breaks. Furthermore, gold nanoparticles (GNPs) are more effective options than gadolinium nanoparticles (GdNPs), and silver nanoparticles (SNPs) for inhibiting and controlling cancer cells.
Zohreh Gholamzadeh; Atieh JozVaziri
Abstract
Thorium is more abundant in nature than uranium. The fertile thorium fuel can breed to fissile U-233 by absorbing a neutron. The produced fissile has good neutronic performance in both thermal and fast neutron spectra. Many types of thorium-based fuels were applied ...
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Thorium is more abundant in nature than uranium. The fertile thorium fuel can breed to fissile U-233 by absorbing a neutron. The produced fissile has good neutronic performance in both thermal and fast neutron spectra. Many types of thorium-based fuels were applied in different nuclear reactors. Also natural thorium oxide is used as seed/blanket configuration that the ThO2 rods are used in the outer sections of any fuel assembly. The present study aims to investigate the ThO2 fuel rod loading in 3000 MW VVER-1000 power reactor. MCNPX and ORIGEN codes were used to evaluate its effects on the core neutronic. In addition, the gamma emission rates of ThO2 spent fuel than the UO2 routine fuel of VVER-1000 was investigated. The obtained results of the computational study showed the ThO2 fuel rod loading in some VVER-1000 fuel assemblies would not end to a breeding behavior of the reactor core even after one-year burnup at 3000 MW power. However, the enriched uranium fuel loading reduction may make a motivation for thorium fuel application in the power reactor.
Akbar Abdi Saray; Hossein Zaki Dizaji; Mortaza Taheri Balanoji
Abstract
To monitor personal safety in the fields of biomedical and health physics, it is necessary to be aware of radiation doses to protect the health and safety of persons. Radiation protection quantities such as air kerma, ambient dose equivalent, and exposure dose rate are obtained by the measured spectrum ...
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To monitor personal safety in the fields of biomedical and health physics, it is necessary to be aware of radiation doses to protect the health and safety of persons. Radiation protection quantities such as air kerma, ambient dose equivalent, and exposure dose rate are obtained by the measured spectrum to determine energy-dependent conversion coefficients/factors. This study aims to obtain and compare an ambient dose equivalent to H∗(10) from the measured gamma-ray spectra by the NaI(Tl) scintillation detector using two various methods. The first method, which is based on the detector response function to find the conversion function, is called the G(E) method. The second method is subdividing the measured gamma-ray spectra into the multiple energy bins, and then obtaining the ambient dose equivalent by using conversion coefficient functions (ω(E)), which were determined by the conversion coefficients (ωi) of each energy bin for three energy intervals of ≤185 keV, 185 to 850 keV, and ≥850 keV. To calculate the detector response matrix and the conversion coefficients of each region of energy, the Monte Carlo simulation code was used for the quasi-mono energetic gamma radiation sources and the synthetic spectra. The results indicate that using the technique based on subdividing the measured spectrum into multiple energy bins helps to avoid the inverse detector response matrix dimension limitations that occur in the G(E) method and also have a lower error percentage in the dose quantity calculation. Consequently, NaI(Tl) scintillation detector has an excellent potential to replace the classical dose rate instruments, i.e. Geiger-Muller, for the early warning of environmental radiation monitoring.
Babak Khanbabaei
Abstract
Indirect drive inertial confinement fusion (ICF) holds promise for achieving practical energy generation through controlled fusion reactions. However, the efficiency of ICF is constrained by the Be ablator material used to contain the fuel. To overcome this limitation, researchers have proposed doping ...
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Indirect drive inertial confinement fusion (ICF) holds promise for achieving practical energy generation through controlled fusion reactions. However, the efficiency of ICF is constrained by the Be ablator material used to contain the fuel. To overcome this limitation, researchers have proposed doping Be with various elements. In this study, we investigate the effects of Na and Br dopants, incorporated at concentrations of 4.86% and 2.1%, respectively, using a one-dimensional MULTI-IFE hydrodynamic code. This code serves as a numerical tool dedicated to analyzing Inertial Fusion Energy microcapsules, facilitating the examination of the Be ablator's performance in indirect drive ICF. Our results indicate that the addition of a beryllium layer doped with Na and Br significantly enhances the target gain, elevating it from the break-even value (G ≈ 1) to approximately G ≈ 12. Furthermore, we delve into the impact of these dopants on the plasma fuel conditions during the implosion, shedding light on the underlying physics of the system. These findings demonstrate that Na and Br doping in the Be ablator represents a viable approach for improving the efficiency of indirect drive ICF, potentially paving the way for the development of practical fusion energy systems.
Javad Karimi; Mohsen Shayesteh; Mehdi Zangian
Abstract
Today, small modular reactors have received considerable attention in various countries. The ABV reactor is a PWR small modular reactor that has various applications. This reactor has been used silumin metal fuel with a 16.5% enrichment. In the present work, the efficiency of the conventional UO2 fuel ...
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Today, small modular reactors have received considerable attention in various countries. The ABV reactor is a PWR small modular reactor that has various applications. This reactor has been used silumin metal fuel with a 16.5% enrichment. In the present work, the efficiency of the conventional UO2 fuel with enrichment of less than 10% to be used as the main fuel of ABV reactor has been investigated, and four different patterns for the reactor core have been proposed. To perform the calculations, the ABV reactor is modeled using the PARCS neutronic code and the RELAP5 thermohydraulic code. Finally, using computational codes for the proposed patterns of the reactor core, various quantities including reactor cycle length, reactivity, burnup, power distribution, fuel, coolant temperature distribution, and feedback coefficients have been calculated.
Omid Safarzadeh; Farahnaz Saadatian-Derakhshandeh
Abstract
The effective β-fraction has a key role in the dynamic response of the reactor. This study aims to assess the suitability and accuracy of the detailed models of DRAGON5 and DONJON5 code for estimation of the effective fraction of delayed neutron for the VVER-1000 reactor core. DRAGON5 is adopted ...
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The effective β-fraction has a key role in the dynamic response of the reactor. This study aims to assess the suitability and accuracy of the detailed models of DRAGON5 and DONJON5 code for estimation of the effective fraction of delayed neutron for the VVER-1000 reactor core. DRAGON5 is adopted to homogenize and condense lattice physics constants of fuel assemblies during fuel burnup, followed by DONJON5, which is used to calculate forward and adjoint flux profiles on the reactor core geometry. A thermal-hydraulic subroutine is developed for VVER-1000 reactor hollow fuel pellets to embody the reactivity feedback raised by changing the reactor power profile. The effective β-fraction is evaluated for each fissile and fertile isotopes in terms of fuel burnup. The results of the coupling scheme are evaluated using the KASKAD code package of Bushehr NPP-I (BNPP-I). The results indicate that the use of SHI and SYBILT modules of DRAGON5 are essential to achieve reasonably precise resolution.
Nahid Hajiloo
Abstract
In this work, the impact of magnetic field presence on the central axis depth-dose curves of helium ion beams inside a heterogeneous phantom with air and bone layers was investigated. According to the calculations, presence of the magnetic field has a considerable impact on the dose distribution of helium ...
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In this work, the impact of magnetic field presence on the central axis depth-dose curves of helium ion beams inside a heterogeneous phantom with air and bone layers was investigated. According to the calculations, presence of the magnetic field has a considerable impact on the dose distribution of helium beams depending on the field strength and beam energy. A 32.3% abrupt increase and 92.5% reduction in dose were observed at the boundary between the water-air and the water-bone layer insert, respectively. The accuracy of the simulation was evaluated by verifying the depth dose curves of helium ion beams in a water phantom with experimental data.
Ali Kolali; ِDavod Naghavi Dizaji; Naser Vosoughi
Abstract
In this study, after discretization of the neutron diffusion equation and adjoint with high-order nodal expansion method in two dimensions and two energy groups, calculations with Momentum and Galerkin weighting functions for rectangular geometry (BIBLIS-2D) and hexagonal geometry (IAEA-2D) reactors ...
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In this study, after discretization of the neutron diffusion equation and adjoint with high-order nodal expansion method in two dimensions and two energy groups, calculations with Momentum and Galerkin weighting functions for rectangular geometry (BIBLIS-2D) and hexagonal geometry (IAEA-2D) reactors are performed. The mean of relative power error for Momentum and Galerkin weighting functions was calculated in BIBLIS-2D reactor 0.42% and 0.62%, respectively, and for IAEA-2D reactor 4.96% and 3.52%, respectively. Regarding the results, it was concluded that in order to increase accuracy with the acceptable time of computing (4 Seconds for rectangular geometry and 28 seconds for hexagonal geometry with Intel® Core™ i7-4510U Processor), the Momentum weighting function for rectangular geometry and the Galerkin weighting function for hexagonal geometry can be used to discretize equations without reducing the node size. Therefore, to increase the accuracy while maintaining the speed of calculations, without reducing the size of nodes, the appropriate weighting function can be used in discretization, which can be very useful in performing calculations of different transients.
Farrokh Khoshahval; Mohammad Rajaee; Nafiseh Tehrani
Abstract
Actinide concentration and activity analysis of the nuclides resulted from the burnup (depletion) process during nuclear reactor operation lifetime is an essential problem in reactor design. Inventory and the corresponding activities of the Tehran Research Reactor (TRR) are evaluated using different ...
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Actinide concentration and activity analysis of the nuclides resulted from the burnup (depletion) process during nuclear reactor operation lifetime is an essential problem in reactor design. Inventory and the corresponding activities of the Tehran Research Reactor (TRR) are evaluated using different methods and compared with each other. WIMS-CITATION, ORIGEN, and MCNP codes are used for plate type inventory calculations. The important actinides, fission products, and fissile inventory ratio of TRR have been calculated at different burnup steps. It is worth noting to mention that knowing the value of inventory helps us for safe handling of the spent fuels and to have a perfect design for transport cask of spent fuels. In this paper, the fuel isotope inventories were calculated for the first and 83rd core configuration of the Tehran Research Reactor, which is named “Core01” and “Core83” respectively. The calculations were first performed using WIMS-D5 and CITATION neutronic codes and then the results are compared with that of ORIGEN and MCNPX code. The total radioactivity of the TRR core at the end of the reactor core life (Core83) is estimated to be 6.47×105 Ci.
Mohsen Mirhabibi; Maryam Najibzadeh; Ali Negarestani; Ahmad Akhound
Abstract
Addressed herein, a new monitoring method for alpha surface contamination based on the function of a thick gaseous electron multiplier (THGEM) in a self-quenching streamer mode (SQS) has been introduced. SQS mode detectors can detect alpha surface contamination in two dimensions. In the current study, ...
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Addressed herein, a new monitoring method for alpha surface contamination based on the function of a thick gaseous electron multiplier (THGEM) in a self-quenching streamer mode (SQS) has been introduced. SQS mode detectors can detect alpha surface contamination in two dimensions. In the current study, the ability of thick gas electron multiplier detectors in SQS mode for two-dimensional monitoring of alpha surface contamination has been investigated by two Am-241 sources with activities equal to 33 and 150 kBq.m-3. It has been found that the brightness is stronger in front of stronger sources. This may be attributed to the difference in contamination levels. It was also observed that the spatial resolution of the contamination rate depends on the number of holes per unit area of each THGEM. The advantage of this system is the ability to determine both the location and intensity of surface contamination with no need for an electronic multiplier or reader system.
