• 시스템엔지니어링학술지
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• 제15권2호
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• pp.39-46
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• 2019

The purpose of this paper is to present the results of the study, applying Systems Engineering(SE) method, on the feasibility of developing a Melt-down Proof Modular Micro Reactor(MDP-MMR) for its future deployment in Korea. The reactor is being developed by NCSU (North Carolina State University) due to its advantage of melt-down proof nature of the reactor core. For this paper, the characteristics of the MDP-MMR has been studied in terms of fuel characteristics, inherent safety features and passive safety system. The NCSU's development process has been reviewed applying the SE method, and further research is recommended for the feasibility study on deploying such a modular micro reactor in Korea.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2000년도 추계학술대회 논문집 학회본부 D
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• pp.638-640
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• 2000

Reactor Noise is defined as the fluctuations of measured instrumentation signals during full-power operation of reactor which have informations on reactor system dynamics such as neutron kinetics, thermal-hydraulics, and structural dynamics. Reactor noise analyses of ex-core neutron detector internals such as fuel assembly and Core Support Barrel in Nuclear Power Plant. A real time mode separation technique have been developed and applied for the analyses. The analyses data base have been constructed for the continuous monitoring and diagnose of the reactor internals. Detailed design of diagnostic system reactor internal structures using neutron noise(RIDS).

• Nuclear Engineering and Technology
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• 제52권11호
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• pp.2460-2470
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• 2020

To study the thermophysical and neutronic properties of thorium-plutonium fuel, a conceptual design of a hybrid facility consisting of a subcritical Th-Pu reactor core and a source of additional D-D neutrons that places on the axis of the core is proposed. The source of such neutrons is a column of high-temperature plasma held in a long magnetic trap for D-D fusionreactions. This article presents computer simulation results of generation of thermonuclear neutrons in the plasma, facility neutronic properties and the evolution of a fuel nuclide composition in the reactor core. Simulations were performed for an axis-symmetric radially profiled reactor core consisting of zones with various nuclear fuel composition. Such reactor core containing a continuously operating stationary D-D neutron source with a yield intensity of Y = 2 × 10¹⁶ neutrons per second can operate as a nuclear hybrid system at its effective coefficient of neutron multiplication 0.95-0.99. Options are proposed for optimizing plasma parameters to increase the neutron yield in order to compensate the effective multiplication factor decreasing and plant power in a long operating cycle (3000-day duration). The obtained simulation results demonstrate the possibility of organizing the stable operation of the proposed hybrid 'fusion-fission' facility.

• 한국전산구조공학회논문집
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• 제27권3호
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• pp.191-198
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• 2014

본 논문에서는 노심용융사고 시 관통노즐이 제거된 원자로용기 하부헤드의 구조 건전성 평가를 수행하였다. 열응력, 노심용융물의 질량 그리고 내압조건의 해석결과를 고려할 때, 하부헤드의 열응력에 의한 영향이 가장 크게 나타났다. 손상 가능성은 파손기준에 따라 평가하였으며, 등가소성변형률이 임계변형률 파손기준보다 낮은 수준으로 평가되었다. 열-구조물 연성해석 결과 하부헤드의 두께 중간층에서 항복강도보다 낮은 응력이 발생한 탄성영역 구간을 확인하였다. 내압이 커지면서 탄성영역 범위가 점차 좁아지면서 탄성영역이 내벽으로 이동하는 결과를 확인하였고, 노심용융사고 시 구조적 건전성을 만족하는 것으로 평가되었다.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2004년도 Proceedings of the 4th Korea-China Joint Workshop on Nuclear Waste Management
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• pp.114-129
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• 2004

The 3D simulations for the Rotary Specimen Rack (RSR), the shielding concret, and the reactor core dismantling processes in the Korea Research Reactor-1&2(KRR-1&2) were carried out in the present work. The four main dismantling items, which are the RSR, reactor core, beam tube, and the thermal column and the shield concrete, were selected among the many components in the KRR-2 by consideration of the activation, worker training, difficulty of the work and so on. On the basis of these, we built 3D CAD models, selected the proper dismantling technologies, and reviewed their dismantling processes. In this study, the 3D simulation results of the shielding concrete, and the reactor core dismantling processes are also presented and discussed.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3127-3132
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• 2021

Feasibility study of burn-up analysis and monitoring using delayed fast neutrons was investigated at Missouri University of Science and Technology Reactor (MSTR). Burnt and fresh fuel elements were used to collect delayed fast neutron data for different power levels. Total reactivity varied depending on the burn-up rate of fuel elements for each core configuration. The regulating rod worth was 2.07E-04 𝚫k/k/in and 1.95E-04 𝚫k/k/in for T121 and T122 core configurations at 11 inch, respectively. Delayed fast neutron spectrum of F1 (burnt) and F16 (fresh) fuel elements were analyzed further, and a strong correlation was observed between delayed fast neutron emission and burn-up. According to the analyzed peaks in burnt and fresh fuels, reactor power dependency was observed and it was determined that delayed neutron provided more reliable results at reactor powers of 50 kW and above.

• Nuclear Engineering and Technology
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• 제51권7호
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• pp.1776-1783
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• 2019

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.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2005년도 추계 학술대회논문집
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• pp.39-44
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• 2005

In HANARO, a multi-purpose research reactor of 30 MWth, the emergency water supply system consists essentially of an emergency water storage tank located in the level of about thirteen meter (13 m) above the reactor core, a three inch ('3\%') diameter water injection pipe line including injection valves from the tank to the reactor cooling inlet pipe and a test loop to do periodic system performance test. When the water level of the reactor pool comes down to the extremely low due to a loss of reactor pool water accident the emergency water stored in the tank should be fed to the core by the gravity force and at that time the design flow rate is eleven point four kilogram per second (11.4 kg/s). But it is impossible periodically to measure the injection flow rate under the emergency condition because the normal water level should be maintained during the reactor operation. This paper describes a flow network analysis to simulate the flow rate under the emergency condition. As results, it was confirmed through the analysis results that the calculated flow rate agrees with the design requirement under the emergency condition.

• Nuclear Engineering and Technology
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• 제50권7호
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• pp.1017-1023
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• 2018

The measurement and calculation of neutronic parameters in nuclear research reactors has an important influence on control and safety of the nuclear reactor. The power peaking factors, reactivity coefficients and kinetic parameters are the most important neutronic parameter for determining the state of the reactor. The position of the control shim safety rods in the core configuration affects these parameters. The main purpose of this work is to use the MTR_PC package to evaluate the effect of the partially insertion of the control rod on the neutronic parameters at the operating core of the Tehran Research Reactor. The simulation results show that by increasing the insertion of control rods (bank) in the core, the absolute values of power peaking factor, reactivity coefficients and effective delayed neutron fraction increased and only prompt neutron life time decreased. In addition, the results show that the changes of moderator temperature coefficients value versus the control rods positions are very significant. The average value of moderator temperature coefficients increase about 98% in the range of 0-70% insertion of control rods.

• Nuclear Engineering and Technology
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• 제48권3호
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• pp.635-641
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• 2016

Because of prohibitive data storage requirements in large-scale simulations, the memory problem is an obstacle for Monte Carlo (MC) codes in accomplishing pin-wise three-dimensional (3D) full-core calculations, particularly for whole-core depletion analyses. Various kinds of data are evaluated and quantificational total memory requirements are analyzed based on the Reactor Monte Carlo (RMC) code, showing that tally data, material data, and isotope densities in depletion are three major parts of memory storage. The domain decomposition method is investigated as a means of saving memory, by dividing spatial geometry into domains that are simulated separately by parallel processors. For the validity of particle tracking during transport simulations, particles need to be communicated between domains. In consideration of efficiency, an asynchronous particle communication algorithm is designed and implemented. Furthermore, we couple the domain decomposition method with MC burnup process, under a strategy of utilizing consistent domain partition in both transport and depletion modules. A numerical test of 3D full-core burnup calculations is carried out, indicating that the RMC code, with the domain decomposition method, is capable of pin-wise full-core burnup calculations with millions of depletion regions.