4- Nuclear Reactor technology
Sarah Kamalpour; Hossein Khalafi
Abstract
The neutronic safety parameters determine reactor dynamic response. These parameters change as a function of core inventory during reactor cycle. Therefore, assessment of reactor behavior throughout operational cycle is an important issue in reactor safety analysis for transients. The purpose of the ...
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The neutronic safety parameters determine reactor dynamic response. These parameters change as a function of core inventory during reactor cycle. Therefore, assessment of reactor behavior throughout operational cycle is an important issue in reactor safety analysis for transients. The purpose of the present study is to evaluate SMART reactor response to changes in neutronic safety parameters during reactor cycle in reactivity insertion accident condition. MCNPX 2.6 nuclear code is utilized to calculate neutronic safety parameters throughout reactor cycle. The reactor dynamic model is simulated based on point kinetic equations and lumped temperatures to predict reactor response to ramp reactivity insertion. Based on this approach, the effect of neutronic parameters on reactor behavior are investigated in the beginning and end of cycle under reactivity transients. Hot full and zero power operational reactor states are considered in the analysis. Results illustrate that reactor at end of cycle has faster response with smaller transient power peak during reactivity insertion accident compared to beginning of cycle. The neutronic parameters, specifically negative feedbacks at both beginning and end of reactor cycle guarantee the safe performance of reactor at all examined conditions. The detailed comparative results are explained in the paper.
1- Experimental and Theoretical Nuclear Physics
Mahdi Eshghi; Sasan Soudi; Sameer M. Khadir
Abstract
The scattering problems, in the presence of an external potential field, have become highly interesting topics in relativistic and non-relativistic quantum mechanics. It’s well known that the scattering of a relativistic particle in the field of a potential can be treated exactly by finding the ...
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The scattering problems, in the presence of an external potential field, have become highly interesting topics in relativistic and non-relativistic quantum mechanics. It’s well known that the scattering of a relativistic particle in the field of a potential can be treated exactly by finding the continuum solutions of the Dirac equation. In this research, we obtain the exact solution to the Dirac equation with the Pöschl-Teller double ring-shaped Coulomb (PTDRSC) potential for any spin-orbit quantum number k. The relativistic scattering amplitude for spin 1/2 particles in the field of this potential has been studied. The wave functions are being expressed in terms of the hyper-geometric series of the continuous states on the k/2pi scale. In addition, a formula for the phase shifts has also been found. In the nonrelativistic limits, our solution to the Dirac particle converges to that of the Schrödinger one. At the high temperature, the partition function is being calculated in order to study the behavior of some thermodynamic properties.
4- Nuclear Reactor technology
Danial Salehi; Gholamreza Jahanfarnia; Ehsan Zarifi
Abstract
Canadian GEN IV Super Critical Water Reactor (Canadian-SCWR) is a combination version of conventional CANDU reactor with the using super critical water as coolant. Thermal-hydraulic analysis of a nuclear reactor is done to ensure that reactor will work in its safety margins. In this study, thermal hydraulic ...
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Canadian GEN IV Super Critical Water Reactor (Canadian-SCWR) is a combination version of conventional CANDU reactor with the using super critical water as coolant. Thermal-hydraulic analysis of a nuclear reactor is done to ensure that reactor will work in its safety margins. In this study, thermal hydraulic analysis of Canadian-SCWR is conducted by numerically solving of conservation equations by a porous media approach. The latest concept of Canadian-SCWR core was used for this purpose. In this concept, in each fuel bundles, super critical water flows in two pass and low pressure and low temperature heavy water moderator flows around fuel channel in the Calandria vessel, separately. Average axial temperature, density, heat capacity, pressure and velocity of supercritical water was estimated in two regions of fuel channels (two pass) i.e centeral flow tubes and the fuel rods channel. Compared to the literature, there is a good agreement between our results and the reported results.
3- Measurement of Radiation and Radioactivity
Mahdieh Sadat Mousavi; Faezeh Rahmani
Abstract
Boron Neutron Capture Therapy (BNCT) is one of the promising treatment methods for some malignant tumors such as Glioblastoma Multiforme (GBM). One of the requirements of BNCT treatment is the accurate and real-time boron-concentration monitoring to ensure the efficacy of treatment and no leakage of ...
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Boron Neutron Capture Therapy (BNCT) is one of the promising treatment methods for some malignant tumors such as Glioblastoma Multiforme (GBM). One of the requirements of BNCT treatment is the accurate and real-time boron-concentration monitoring to ensure the efficacy of treatment and no leakage of boron. An accurate method for real-time calculation of the boron dose distribution mapping during irradiation is Single Photon Emission Computed Tomography (SPECT), in which the determination of boron distribution is based on the detection of 478 keV prompt gamma-rays generated through thermal neutron capturing by B-10. Tehran Research Reactor (TRR) is the only possible source for BNCT in Iran, so as the first approach, this BNCT-SPECT system has been evaluated for TRR. In this paper, an imaging system of BNCT-SPECT including four arrays of collimator/detector has been designed for real-time dosimetry as well as boron-10 concentration distribution map in the phantom that placed in front of the therapeutic neutron beam using the Monte Carlo simulation code MCNP6. Maximum Likelihood Expectation Maximization (MLEM) method has been used for image reconstruction which results 1 cm spatial resolution.
4- Nuclear Reactor technology
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.
