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- 저자 : Mohammad Hossein Esteki (검색결과 2 건)
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Safavi Amir,Esteki Mohammad Hossein,Mirvakili Seyed Mohammad,Arani Mehdi Khaki 한국원자력학회 2020 Nuclear Engineering and Technology Vol.52 No.8
Due to ever-growing advancements in computers and relatively easy access to them, many efforts have been made to develop high-fidelity, high-performance, multi-physics tools, which play a crucial role in the design and operation of nuclear reactors. For this purpose in this study, the neutronic Monte Carlo and thermal-hydraulic sub-channel codes entitled MCNP and COBRA-EN, respectively, were applied for external coupling with each other. The coupled code was validated by code-to-code comparison with the internal couplings between MCNP5 and SUBCHANFLOW as well as MCNP6 and CTF. The simulation re-sults of all code systems were in good agreement with each other. Then, as the second problem, the core of the VVER-1000 v446 reactor was simulated by the MCNP4C/COBRA-EN coupled code to measure the capability of the developed code to calculate the neutronic and thermohydraulic parameters of real and industrial cases. The simulation results of VVER-1000 core were compared with FSAR and another nu-merical solution of this benchmark. The obtained results showed that the ability of the MCNP4C/COBRA-EN code for estimating the neutronic and thermohydraulic parameters was very satisfactory.
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Ali Farsoon Pilehvar,Mohammad Hossein Esteki,Afshin Hedayat,GHOLAM REZA ANSARIFAR 한국원자력학회 2018 Nuclear Engineering and Technology Vol.50 No.5
Self-pressurization analysis of the natural circulation integral nuclear reactor through a new dynamicmodel is studied. Unlike conventional pressurized water reactors, this reactor type controls the systempressure using saturated coolant water in the steam dome at the top of the pressure vessel. Selfpressurizationmodel is developed based on conservation of mass, volume, and energy by predictingthe condensation that occurs in the steam dome and the flashing inside the chimney using the partialdifferential equation. A simple but functional model is adopted for the steam generator. The obtainedresults indicate that the variable measurement is consistent with design data and that this new model isable to predict the dynamics of the reactor in different situations. It is revealed that flashing andcondensation power are in direct relation with the stability of the system pressure, without whichpressure convergence cannot be established.
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