1,721,316 research outputs found
미포화 유동 비등에서 가시화를 통한 가열면 근처 기포 거동에 관한 연구
No full text학위논문(석사) - 한국과학기술원 : 원자력공학과, 2000.2, [ vi, 54 p. ]The behavior of near-wall bubbles in subcooled flow boiling has been investigated photographically to identify the physical mechanisms of critical heat flux at subcooled and low-quality conditions. Visualization experiments were performed for water flow in vertical rectangular channels under atmospheric pressure for mass fluxes below 2020 kg/㎡s. The thickness and other features of the near-wall bubble layer were examined with the aid of a high-speed camera, a still camera and an 8 mm-camera recorder. The number of activated nucleation sites increased as the wall heat flux was increased. At sufficiently high heat flux about 5-7.5 MW/㎡, the appearance of vapor clot or blanket on the heated surface made a role of an obstacle between main liquid region and the region near heater. At such high heat flux, three characteristic regions were observed in the heated channel: (a) a superheated liquid layer with attached bubbles, (b) a flowing bubble layer consisting of large coalesced bubbles over the attached bubbles, and finally (c) the liquid core.한국과학기술원 : 원자력공학과
비등 위기에서 액체 미세층 현상의 직접 관찰 및 나노 유체 성능에 관한 연구
No full text학위논문(박사) - 한국과학기술원 : 원자력및양자공학과, 2004.8, [ xv, 222 p. ]The boiling crisis phenomenon is the most enormously studied and disputed topic in the area of boiling heat transfer. The great interest is due to practical motives since it is desirable to design a heat exchanger or boiler to operate at as high a heat flux as is possible with optimum heat transfer rates, but without the risk of physical burnout. This study consists of two parts of boiling crisis mechanism-related visualization study and CI-1F enhancement-related nano-fluid study. In Part I, firstly subcooled flow boiling phenomena have been investigated photographically for water in vertical, one-side heated, rectangular channels under atmospheric pressure. At sufficiently high heat fluxes, three characteristic layers were observed in the heated channel: (a) a superheated liquid layer with small bubbles attached on the heated wall, (b) a flowing bubble layer consisting of large coalesced bubbles over the superheated liquid layer, and (c) the liquid core over the flowing bubble layer. In addition, the existence of a liquid sublayer under coalesced bubbles was identified photographically. According to visualization, the CI-1F mechanism for the present experimental condition could be related to the formation of large vapor clots resulting from coalescences of bubbles and the evaporation of the superheated liquid layer beneath those clots. Secondly, a photographic study of subcooled flow boiling with R-134a has been performed under higher pressure. The visualization made possible a detailed observation of the characteristic near-wall region, consisting of vapor remnants, an interleaved liquid layer, and coalesced adjacent bubbles. In addition, it is shown that the near-wall bubble layer of nucleate boiling beneath vapor clots is extinguished and, afterwards, the heated surface is locally covered by large vapor films, at CHF. Thirdly, a visualization study of pool boiling with nano-fluids has been performed. In the boiling, a thin liquid film adhering to a heater s...한국과학기술원 : 원자력및양자공학과
AN AXIOMATIC DESIGN APPROACH OF NANOFLUID-ENGINEERED NUCLEAR SAFETY FEATURES FOR GENERATION III plus REACTORS
No full textA variety of Generation III/III+ reactor designs featuring enhanced safety and improved economics are being proposed by nuclear power industries around the world to solve the future energy supply shortfall. Nanofluid coolants showing an improved thermal performance are being considered as a new key technology to secure nuclear safety and economics. However, it should be noted that there is a lack of comprehensible design works to apply nanofluids to Generation III+ reactor designs. In this work, the review of accident scenarios that consider expected nanofluid mechanisms is carried out to seek detailed application spots. The Axiomatic Design (AD) theory is then applied to systemize the design of nanofluid-engineered nuclear safety systems such as Emergency Core Cooling System (ECCS) and External Reactor Vessel Cooling System (ERVCS). The various couplings between Gen-III/III+ nuclear safety features and nanofluids are investigated and they try to be reduced from the perspective of the AD in terms of prevention/mitigation of severe accidents. This study contributes to the establishment of a standard communication protocol in the design of nanofluid-engineered nuclear safety systems.This work has been supported by the research start-up fund of the UNIST faculty (grant number 1.090011)and Nuclear Research & Development Program throughthe National Research Foundation of Korea (NRF) fundedby the Ministry of Education, Science and Technology(grant number 20090078277)
PIV Flow Measurement in the Moderator Circulation Test (MCT) Facility under Isothermal Condition
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Adjoint-based sensitivity analysis of circulating liquid fuel system for the multiphysics model of molten salt reactor
No full textThe strongly coupled behaviors between neutronics and thermal-hydraulics of liquid-fueled molten salt reactors make it difficult to evaluate system behaviors, due to the transport of precursors along moving fuel. Extending an adjoint-based method on the multiphysics approach, different assumptions on temperature dependencies of nuclear and thermophysical properties of salt are included in the local sensitivity analysis of a circulating liquid fuel system. Local sensitivity of various types of system response in steady-state is analyzed for 39 parameters including coupling models, reactor design values, and kinetic constants of delayed neutron and decay heat precursors for a simplified 1D model of molten salt fast reactor. Extended adjoint-based sensitivity analysis method for MSR is successfully validated achieving 1.38% deviation on average between a recalculation and adjoint method, comparing local sensitivities to all parameters. Also, it takes 66.3 times less in computational time compared with the recalculation method for evaluating the sensitivity of the same type of system response. The importance of all the parameters to the system response is analyzed according to the assumptions on temperature dependencies to nuclear data and salt properties. The most influencing ones are fission energy-related terms, and their importance increases when temperature dependencies are taken into account, compared with constant properties. Changes of influences on the sensitivity are investigated from the relative changes of the parameter values in various system response types, and it implies the importance to consider the multiphysics modeling on the local sensitivity analysis
Design and Performance of Hybrid Control Rod For Passive IN-core Cooling System
Get PDFDepartment of Nuclear EngineeringProtection of the public and the environment from undue radiation hazards is a definition of nuclear safety. Although there are various safety systems in nuclear power plants to achieve the nuclear safety, Fukushima-Daiichi accident showed the vulnerabilities of the installed safety systems. After the Fukushima accident, various passive safety systems and strategies are under development to cope with the postulated accidents. The majority of passive safety systems concentrated to inject emergency core coolant (ECC) or feedwater with the circuits comprise many pipelines and valves. In station blackout condition, the pressure of reactor vessel would be higher than the ECC injection pressure resulting in failure of ECC supply and eventually causing core damage. The reliability issues about the performance of passive safety systems have been discussed owing to their high uncertainties, low performance, and lack of experience in operation compared to active safety systems. In aspect of probabilistic safety, complex circuits which comprise many valves and pipelines have possibilities of single failure and common cause failure. Development of innovative passive safety system having differentiated working principle, significant performance, and low possibility of failure can enhance reactor safety providing solutions for the aforementioned problems. Based on these requirements, hybrid control rod which combines the functions of control rod and heat pipe was proposed for the development of passive in-core cooling system (PINCs). The control rods drop to the core using gravity and shutdown the reactor by neutron absorption. The thermosyphon heat pipe is a passive heat transfer device using phase change and convection of working fluid in a closed metal container having two different temperature interfaces (evaporator and condenser). The combination of thermosyphon and control rod, hybrid control rod can achieve reactor shutdown and decay heat removal simultaneously at accident conditions.
Hybrid control rod was designed considering the aspects of neutronics (reactivity worth) and mechanical integrity. Most of the nuclear reactors operate at high temperature and high pressure environment with high power density. Thus, pressure control strategies of the hybrid control rod using non-condensable gas and expansion of the working fluid were established to achieve high decay heat removal capacity and operating conditions. The designed hybrid control rods were equipped on the experimental facility and their thermal performances were studied under various amount of working fluid, amount of non-condensable gas, and operating pressures of the test section. The experimental results showed relations between heat transfer characteristics and controlled parameters. Controlling operating condition of hybrid control rod in high pressure worked successfully, and the proportionality between maximum heat removal capacity and operating pressure of hybrid control rod design has been proven. Measured maximum heat transfer rate of single hybrid control rod was 6 kW at 20 bar.
Simulations of multi-dimensional analysis for reactor safety (MARS) code were also performed to validate the experimental results and evaluate the prediction capability of the code on the hybrid control rod. The simulation results showed the limits of heat transfer models in the code analyzing the hybrid control rod in which the boiling and condensation heat transfer occurs simultaneously in a manner of countercurrent flow. The experimental results were compared with several models associated with boiling heat transfer, condensation heat transfer, and critical heat flux (CHF) of thermosyphon for the development or the selection of optimal models. The selected models could be implemented to system analysis codes in the purpose of deterministic safety assessment of PINCs against design basis accidents. Imura???s correlation, which was developed in two-phase natural convection condition and validated with experiments in wide range, was selected as boiling heat transfer model of pressurized hybrid control rod. The existing condensation models were based on Nusselt???s film condensation theory. Hence, the effect of non-condensable gas and perturbation between upward vapor flow and downward liquid film flow were not considered at the same time. The change of effective heat transfer length due to presence of non-condensable gas and effect of fluid inertia were considered for the derivation of new condensation model. The main thermal-hydraulic phenomenon which induces CHF of thermosyphon is flooding. The flooding-based CHF models for thermosyphon were derived with theories on instability of the liquid film or maximum liquid film flow rate in countercurrent flow condition. The limited prediction capabilities of the models were attributed to difference between hydraulic diameter and heated diameter as well as high operating pressure. Consequently, new model regarding the CHF of hybrid control rod was suggested to explain its unique characteristics.
The hybrid control rod could be equipped on spent fuel dry storage casks for the extension of their thermal margins. The mock-up was designed to be scaled-down to 1/10 of metal dry storage cask developed by NAC. The effect of hybrid control rod on thermal margins of the cask was experimentally studied. The equipment of hybrid control rod with installation of heat sink lid reduced the temperature distributions inside the cask at equal power density condition. Application of hybrid control rod could extend the thermal margin up to 30 %. Feasibility of PINCs based on experimentally and analytically studied hybrid control rods were discussed according to commercial reactors.
A number of nuclear facilities has been built to supply and manage energy. The nuclear fuels generate decay heat even in shutdown condition by fission products. Management of the decay heat is important to satisfy demand for nuclear safety. Therefore, new conceptual safety system is required to supplement the issues on existing safety systems. Passive in-core cooling system based on hybrid control rod is the effective way to be applied on extensive nuclear facilities containing nuclear fuels. Pressurized hybrid control rod could meet the operating conditions of application objects with significant decay heat removal capacity.ope
Gallium Passive Decay Heat Removal Systems Design and Evaluation for IVR-ERVCS of APR 1400 and PDHRS of UCFR-100
No full textDepartment of Nuclear EngineeringThe passive decay heat removal system is one of the important concepts of nuclear power plants. Because the principle of the system is based on laws of physics such as gravity or natural convection, it is able to function without electric power or actuation by control equipments. In this paper, the liquid metal, gallium-cooled passive decay heat removal systems of pressurized water reactor (PWR), APR 1400 and Ultra-long-life-Core fast Reactor, UCFR-100 were suggested and simulated using MARS-Ga code. The experimental study to investigate the natural convective heat transfer was also conducted.
Although MARS-LMR code was originally intended for a safety analysis of liquid metal-cooled fast reactor such as a sodium-cooled fast reactor, gallium properties were newly added to this code as so-called MARS-Ga code which is applicable for gallium cooled systems. The implementation of the properties for liquid metals in MARS-LMR code used the soft-sphere model based on Monte Carlo calculations for particles interacting with pair potentials, and the transport properties such as surface tension, thermal conductivity, and dynamic viscosity for the liquid and vapor state were also included.
In the experimental analysis, the natural convection heat transfer of liquid gallium was investigated in the rectangular loop which consists of an indirect heating block test section, a condenser, and the 1/2 inch SS 316L tubes as well as the orifice for measuring mass flowrate. The average Nusselt number of liquid gallium was measured within the heat flux range of 6.17??103 ? 5.07??104 W/m2. The mass flow rate for the natural convection of liquid gallium depending on power level was also compared by using CFD and MARS-Ga code.
In the numerical analyses, the evaluations of the gallium-water in-vessel corium retention through external reactor vessel cooling system (IVR-ERVCS) in APR 1400 and gallium-cooled passive decay heat removal system (PDHRS) in UCFR-100 were performed using MARS-Ga code. The attractive properties such as the low melting point, the high boiling point, and no reaction with water ensure that gallium can play an important role in nuclear safety as an alternative coolant in PWRs and SFRs.
In the gallium-water IVR-ERVCS, the generated decay heat is transferred to liquid gallium through the reactor pressure vessel and then removed from the water pool as a heat sink. The numerical analysis results showed that the temperature range of the liquid gallium is much lower than its boiling point and confirm the natural convection under a medium break loss of coolant accident (MBLOCA) and large break loss of coolant accident (LBLOCA). Because liquid gallium in this system didn???t have a phase change, unlike water, the gallium-water IVR-ERVCS can provide stable and reliable cooling capability. Sensitivity studies were also performed by changing several parameters such as the initial temperature of liquid gallium and water pool inventory, and their results indicated that the working time of the gallium-water IVR-ERVCS depends on the inventory of the water pool using MARS-Ga code.
UCFR-100 is a 260MWth / 100MWe sodium-cooled fast reactor which requires no on-site refueling. UCFR-100 is a pool type reactor including the metallic fuels, intermediate heat exchangers, steam generators, and gallium-cooled PDHRS unlike the existing designs of sodium fast reactors. The safety analysis was performed for loss of flow (LOF) due to the pumping failure of primary pumps using MARS-Ga code. As a result, it confirmed that the liquid gallium can work properly as a boundary material between sodium and atmosphere for steady state and transient situation in UCFR-100.ope
Effects of Graphene and SiC Nanofluids on Critical Heat Flux and Quenching for Advanced Nuclear Reactors
No full textDepartment of Nuclear Engineering Program, PhD thesisMy research purpose is the study on effects of graphene and SiC nanofluids for advanced nuclear reactors through the experiment on flow boiling Critical Heat Flux (CHF) enhancement and the quenching experiment. In here, nanofluids are nanotechnology-based fluids engineered for enhancing thermal conductivity by dispersing and stably suspending nanoparticles in traditional heat transfer fluids. In the present study, two kinds of works were conducted.
First, the CHF is characterized by a sudden reduction of the local Heat Transfer Coefficient (HTC) that results from the replacement of liquid by vapor adjacent to the heat transfer surface and ordinarily, represents the thermal limitation in which a phase change happens during heating. When the CHF occurs, an inordinate decrease in the heat transfer rate for heat flux and temperature controlled system generates. Moreover, it is generally more important in applications such as power generation for heat flux controlled system because of maintenance of the integrity occurring in heated surface. So, it is very important to enhance the CHF to ensure the system safety and improve the efficiency. Many methods to enhance the CHF have been investigated and a new technique in recent years among these methods is nanofluids technology. The influences of 0.01 volume fraction (%) Al2O3, SiC and Graphene Oxide (GO)/water nanofluids and fluid thermal hydraulic conditions on CHF have been experimented. Experiments were performed using 1/2 inch SS 316L tube when the mass flux is 100, 150, 200, 250, 300 kg/m2s and inlet temperature is 25 and 50 ??C. The maximum CHF enhancement of Al2O3/water nanofluid was 15 % at inlet temperature of 50 ??C and mass flux of 200 kg/m2s. That of SiC/water nanofluid was 41 % at inlet temperature of 25 ??C and mass flux of 150 kg/m2s. And, the maximum CHF enhancement of GO/water nanofluid was 100 % at inlet temperature of 25 ??C and mass flux of 250 kg/m2s. The CHF enhancements of nanofluids were caused to enhanced wettability of the liquid film on the heater surface due to the deposition of nanoparticles. The enhanced wettability is due to the change of surface structure (porous structure). This is confirmed through macroscopic observation, SEM observation and contact angle measurement. Liquid film thickness affected by evaporation, entrainment and deposition mass transfer can be closely linked with wettability and nanoparticles properties. Also, the CHF enhancement of nanofluids is caused to increase of thermal activity related to thermal conductivity and thickness.
Second, quenching experiments were conducted to investigate the effect of nanofluids on reflood heat transfer in a long vertical tube (1,300 mm in the heating length). When the potential application of nanofluids comes to Emergency Core Cooling System (ECCS), the situation of interest is quenching phenomena of fuel rods during reflood of emergency coolants. The reflood tests have been performed using SiC and GO/water nanofluids as a coolant, instead of water. We have observed a more enhanced cooling performance in the case of the nanofluids reflood. A cooling performance (quenching time) is enhanced more than 20 seconds for SiC/water and GO/water nanofluids. A more enhanced cooling performance is attributed to a high wettability of a thin layer formed on a heating surface by a deposition of nanoparticles. The enhanced wettability is due to the change of surface structure (porous structure). The enhancing cause of the cooling performance using the nanofluids were investigated through macroscopic observation, SEM, SEM-EDS and contact angles of the inner surface of the test section. Also, a more enhanced cooling performance is caused to increase of thermal activity related to thermal conductivity and thickness.
Effects of graphene/SiC nanofluids show the enhancement of safety margin for advanced nuclear reactors in terms of CHF enhancement and an enhanced quenching performance.ope
Study on the Multiphysics Modeling of Molten Salt Reactor Using Adjoint-based Sensitivity Analysis Method
Full text linkDepartment of Nuclear EngineeringTo pursue the nuclear energy as future energy resources, several reactor types of the nuclear power plant has been developed to achieve the advances in sustainability, reduction of nuclear waste, safety and reliability, proliferation resistance and competitive economies in the framework of the Generation IV International Forum (GIF). Among them, Molten Salt Reactor (MSR) is the only reactor that adopts liquid fuel serving as a coolant as well, which is not a new concept in retrospect of the history of the nuclear reactor. Fermi???s water boiler, the first nuclear reactor fueled with uranium-235 enriched uranium, designed and operated with aqueous homogeneous fuel in liquid form. It adopts liquid fuel having lots of advantages to achieve high power density and inherent safety. However, it also has some technical limitations in maintaining homogeneous state with relatively large fuel particle at that time. With a lesson of previous experiences, it was evolved to adopt the liquid fuel as the eutectic formation of fuel material with molten salt to solve the issue on the slurry fuel, which is the current MSR concept. This was successfully demonstrated by Aircraft Reactor Experiment (ARE) and Molten Salt Reactor Experiment (MSRE) by Oak Ridge National Laboratory in the 1950s. Recently, liquid-fueled MSR is reconsidered as an alternative of a conventional nuclear power plants with aims of the development of related chemical reprocessing and molten salt technologies, taking advantages of the liquid fuel in safety and economic viewpoint as well as the radioactive waste issue. In a viewpoint of design and analysis of liquid-fueled MSR system, complex system behavior should be considered in the aspects of coupled physics; neutronics and thermal-hydraulics. Different from a solid-fueled reactor, delayed neutron precursors generated in core decay at the different location from the fuel flow, and it affects the overall neutron economy. It means a distribution of the fission source is largely dependent on the velocity field which determines the overall power profile. Moreover, it affects temperature field and varies the density of liquid fuel ultimately. Because of this, Multiphysics approach on the MSR system looks promising on the assessment of the system including important aspects of coupled physics. With aims of growing computational power, considering all physics lying under certain problem can be a feasible option nowadays. However, Multiphysics model itself for the complex system should be assessed not to misinterpret the behavior along with system analysis. In the above context, this work is aimed at developing the integrated design and analysis tool for MSR on the Multiphysics approach, which enables to perform system analysis and model sensitivity analysis simultaneously including all underlying physics based on the adjoint method and its application on the assessment of new conceptual design of MSR: nanofluidic molten salt reactor.
The first part explains the adjoint-based sensitivity analysis method on the Multiphysics approach. Adjoint method is the concept to establish a relatively simple problem having duality with a primal system. The concept of the adjoint formulation is independent of the model/input parameter itself, such that it has advantages of the calculating a set of the sensitivity of the certain physical system to the numerous parameters with less computational efforts, compared to traditional way of sensitivity analysis, i.e. recalculating perturbed state. To evaluate adjoint-based sensitivity method for Multiphysics problem, a one-dimensional steady-state model of the circulating liquid fuel system and its sensitivity system is established, which consists of one group neutron diffusion equation, balance equations of 6 groups of neutron precursors and 3 groups of decay heat groups, and energy conservation equation. For the condition of Molten Salt Fast Reactor (MSFR) developed by European project SAMOFAR as a representative of liquid-fueled MSR, physical interpretation of the model sensitivity considering coupled effect of two physics are discussed in terms of modeling option and importance of parameter. The second part includes the development of integrated solver within open source Multiphysics toolkit, OpenFOAM, called msrAdjointFoam. It consists of neutronics and thermal-hydraulics coupled in the same environment (i.e. internal coupling) and model sensitivity analysis solver based on the adjoint formulation of the local sensitivity of system variable to all input/model parameters. Adjoint sensitivity solver is implemented based on the mathematical derivation of model equations of the system. Validation and verification of the solver are conducted with several benchmark cases and compared with the analytic solution. The last part describes the application of the integrated design and analysis tool for developing a new conceptual design of MSR; Nanofluidic Molten Salt Reactor. The concept of nanofluid itself is for enhancement of the convection heat transfer with the adoption of excellent thermal properties with the nanoparticle. To evaluate the conceptual design of the nanofluidic molten salt reactor, msrAdjointFoam was extended to nanoMsrAdjointFoam by implementing nanofluid characteristic; dispersion model suggested by Y. Buongiorno based on the concentration of nanoparticle, and its influences on the coupled neutronics and thermal-hydraulics. Using the integrated analysis tool, several design options of the nanofluidic molten salt reactor including decay heat removal system for drain tank are assessed in terms of system performance and safety.
According to the mathematical background of the concept of adjoint sensitivity system, it can be extended to the model sensitivity analysis of any engineering system that can represent in a PDE form. In addition, the sensitivity analysis method on the Multiphysics approach can give a physical insight in economic and straightforward way. Integrated Multiphysics tool developed can help to understand and evaluate the complex system such as a nuclear reactor in a more realistic way without any exaggeration of the prediction of overall system behavior considering all coupled phenomena. In the end, from the practical point of view, the concept of nanofluidic molten salt reactor is expected to be the most feasible reactor option with enhanced safety, reduction of nuclear waste, high proliferation resistance as a future nuclear power.clos
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