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Mag.etohydrodynamic Calculations of Toroidal Fusion Reactor to Ensure Stable Control

Open Journal of Applied Sciences 2016年第7期6卷 402-408页

作者： Aybaba Hançerlioğullari Asli Kurnaz yosef g. ali madee Department of Physics Faculty of Arts and Sciences Kastamonu University Kastamonu Turkey

The development of mag.etic config.rations to confine the stability fluid plasmas for fusion energ. is a challeng. that is a mixture of basic fusion eng.neering.and invention. In order to keep the fusion reactions in the plasma to be continuing.in the fusion reactors, the speed of tritium breeding.(TBR) should be kept above a certain value. At the Apex fusion reactor, a fast flowing.thin liquid wall has replaced the solid first wall concept of the traditional reactors. Behind the fast flowing.thin liquid wall, a slower and thicker second liquid wall (coat) is present. Monte Carlo Random method (MCRS) is the g.neral name for the solution of experimental and statistical problems with a random approach. This method is dependent upon the theory of probability. In the present work, Mhd impacts are investig.ted quite unimportant for Flibe salt solutions. In this study, the fissile fuel production calculations are done for a neutron wall load of 10 MW/m2 fissile fuel production rates of 238U(n, γ)239Pu and 232Th(n,γ)233U increases almost linearly with increased heavy metal content.

关键词： Fusion Reactor Monte Carlo Mag.etohydrodynamic Tritium Breeding.(TBR)

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Estimates of the Fast and Termal Flux in Blanket of Critical Reactors by Using.Multi-g.oup Methods

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Open Journal of Applied Sciences 2017年第2期7卷 68-81页

作者： Aybaba Hancerliog.llari Asli Kurnaz yosef g.ali madee Ltfei A.Abdalsmd Salem A.A.Shufat Khaled M.Elhadad Hand Hadia Almezog. Mansur Mohamed ali Mansur Faculty of Arts and Sciences Kastamonu UniversityKastamonuTurkey Eng.neering.Faculty Kastamonu UniversityKastamonuTurkey

In this study, based differential equations methods are used to solve equations because these methods are dependent on boundary value data more than other mathematical equations. We have calculated neutron flux, criticality and g.ometrical eig.nvalue by using.multi-g.oup method and solving.the neutron diffusion equation for finite and infinite cylindrical and spherical reactors in this study. For the calculation of the total neutron flux cross sections, we need the neutron diffusion equation. Thus, we have established the relationship between neuron flow and cross-section of neuron depending.on neutron energ.. Critical calculations have been made by comparing.the results with MNCP (montecarlo n-partical) simulation methods. For necessary computer calculations, the prog.amme, Wolfram-Matematica-7 has been used.

关键词： Critical Reactor Neutron Diffusion Equation Mcnp Multi-g.oup Method Simulation

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