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RISS 인기검색어
- 검색키워드
- 저자 : Pirouzmand, Ahmad (검색결과 5 건)
- 무료
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Pirouzmand, Ahmad,Hadad, Kamal,Suh, Kune Y. Korean Nuclear Society 2011 Nuclear Engineering and Technology Vol.43 No.3
This paper considers the concept of analog computing based on a cellular neural network (CNN) paradigm to simulate nuclear reactor dynamics using a time-dependent second order form of the neutron transport equation. Instead of solving nuclear reactor dynamic equations numerically, which is time-consuming and suffers from such weaknesses as vulnerability to transient phenomena, accumulation of round-off errors and floating-point overflows, use is made of a new method based on a cellular neural network. The state-of-the-art shows the CNN as being an alternative solution to the conventional numerical computation method. Indeed CNN is an analog computing paradigm that performs ultra-fast calculations and provides accurate results. In this study use is made of the CNN model to simulate the space-time response of scalar flux distribution in steady state and transient conditions. The CNN model also is used to simulate step perturbation in the core. The accuracy and capability of the CNN model are examined in 2D Cartesian geometry for two fixed source problems, a mini-BWR assembly, and a TWIGL Seed/Blanket problem. We also use the CNN model concurrently for a typical small PWR assembly to simulate the effect of temperature feedback, poisons, and control rods on the scalar flux distribution.
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AHMAD PIROUZMAND,KAMAL HADAD,서균렬 한국원자력학회 2011 Nuclear Engineering and Technology Vol.43 No.3
This paper considers the concept of analog computing based on a cellular neural network (CNN) paradigm to simulate nuclear reactor dynamics using a time-dependent second order form of the neutron transport equation. Instead of solving nuclear reactor dynamic equations numerically, which is time-consuming and suffers from such weaknesses as vulnerability to transient phenomena, accumulation of round-off errors and floating-point overflows, use is made of a new method based on a cellular neural network. The state-of-the-art shows the CNN as being an alternative solution to the conventional numerical computation method. Indeed CNN is an analog computing paradigm that performs ultra-fast calculations and provides accurate results. In this study use is made of the CNN model to simulate the space-time response of scalar flux distribution in steady state and transient conditions. The CNN model also is used to simulate step perturbation in the core. The accuracy and capability of the CNN model are examined in 2D Cartesian geometry for two fixed source problems, a mini-BWR assembly, and a TWIGL Seed/Blanket problem. We also use the CNN model concurrently for a typical small PWR assembly to simulate the effect of temperature feedback, poisons, and control rods on the scalar flux distribution.
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Neutronic assessment of BDBA scenario at the end of Isfahan MNSR core life
M. Ahmadi,A. Pirouzmand,A. Rabiee 한국원자력학회 2018 Nuclear Engineering and Technology Vol.50 No.7
The present study aims to assess the excess induced reactivity in a Miniature Neutron Source Reactor(MNSR) for a Beyond Design Basis Accident (BDBA) scenario. The BDBA scenario as defined in the SafetyAnalysis Report (SAR) of the reactor involves sticking of the control rod and filling of the inner and outerirradiation sites with water. At the end of the MNSR core life, 10.95 cm of Beryllium is added to the top ofthe core as a reflector which affects some neutronic parameters such as effective delayed neutronsfraction (beff), the reactivity worth of inner and outer irradiation sites that are filled with water and thereactivity worth of the control rod. Given those influences and changes, new neutronic calculations arerequired to be able to demonstrate the reactor safety. Therefore, a validated MCNPX model is used tocalculate all neutronic parameters at the end of the reactor core life. The calculations show that theinduced reactivity in the BDBA scenario increases at the end of core life to 7.90 ± 0.01 mk which issignificantly higher than the induced reactivity of 6.80 mk given in the SAR of MNSR for the samescenario but at the beginning of the core's life. Also this value is 3.90 mk higher than the maximumallowable operational limit (i.e. 4.00 mk).
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Effect of test-caused degradation on the unavailability of standby safety components
S. Parsaei,A. Pirouzmand,M.R. Nematollahi,A. Ahmadi,K. Hadad Korean Nuclear Society 2024 Nuclear Engineering and Technology Vol.56 No.2
This paper proposes a safety-critical standby component unavailability model that contains aging effects caused by the elapsed time from installation, component degradation due to surveillance tests, and imperfect maintenance actions. An application of the model to a Motor-Operated Valve and a Motor-Driven Pump involved in the HPIS of a VVER/1000-V446 nuclear power plant is demonstrated and compared with other existing models at component and system levels. In addition, the effects of different unavailability models are reflected in the NPP's risk criterion, i.e., core damage frequency, over five maintenance periods. The results show that, compared with other models that do not simultaneously consider the full effects of degradation and maintenance impacts, the proposed model realistically evaluates the unavailabilities of the safety-related components and the involved systems as a plant age function. Therefore, it can effectively reflect the age-dependent CDF impact of a given testing and maintenance policy in a specified time horizon.
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M. Ahmadi,A. Rabiee,A. Pirouzmand 한국원자력학회 2019 Nuclear Engineering and Technology Vol.51 No.7
To accurately analyze the accidents in nuclear reactors, a thermohydraulic-neutronic coupling calculation is required to solve fluid dynamics and nuclear reactor kinetics equations in fine cells simultaneously and evaluate the local effects of neutronic and thermohydraulic parameters on each other. In the present study, a 3D thermohydraulic-neutronic coupling model is developed, validated and then applied for Isfahan MNSR (Miniature Neutron Source reactor) safety analysis. The proposed model is developed using FLUENT software and user defined functions (UDF) are applied to simulate the neutronic behavior of MNSR. The validation of the proposed model is first evaluated using 1mk reactivity insertion experiment into Isfahan MNSR core. Then, the developed coupling code is applied for a design basis accident (DBA) scenario analysis with the insertion of maximum allowed cold core reactivity of 4 mk. The results show that the proposed model is able to predict the behavior of the reactor core under normal and accident conditions with a good accuracy.
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