• Title/Summary/Keyword: Coolant temperature coefficient

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Transition mechanism during the critical heat flux condition in flow and pool boiling (유동 및 풀비등에 있어서 한계열플럭스 상태하의 천이기구)

  • 김경근;김명환;권형정;김종헌;최순호
    • Journal of Advanced Marine Engineering and Technology
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    • v.13 no.4
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    • pp.40-53
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    • 1989
  • Boiling heat transfer phenomena is widely applied to BWR and electrical heating system because of its high heat transfer coefficient. In these systems, steady state heat transfer is dependent on nucleate boiling. When the heat generating rate is sharply increased or the cooling capacity of coolant is sharply decreased, sharp wall temperature rise is occurred under the critical heat flux(CHF) condition. This paper presents the simple wall temperature fluctuation model of transition mechanism in the repeating process of overheating and quenching, when coalescent bubble passes relatively slowly on the wall and simultaneously the transition from nucleate boiling to film boiling is carried at especially onset of the CHF state. The values calculated by the present model are resulted comparatively good with the measured.

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Studies on the Film Cooling Characteristics of Turbine Blade Cylindrical and Shaped Holes (원통형과 변형된 분사홀을 갖는 터빈 블레이드의 막냉각 특성에 관한 연구)

  • Kim, S.-M.;Kim, Youn J,;Cho, H.-H.
    • 유체기계공업학회:학술대회논문집
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    • 2001.11a
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    • pp.334-338
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    • 2001
  • In order to investigate the effects of various injection hole shapes on the film cooling of turbine blade, three test models having cylindrical and shaped holes were used. A three-dimensional Navier-Stokes code with standard k-$\epsilon$ model was used to compute the film cooling coefficient on the film cooled turbine blade. Over 330,000 grids were used to compute the flow over the blade. Mainstream Reynolds number based on the cylinder diameter was $7.1{\times}10^4$. The turbulence intensity kept at $5.0\%$ for all inlets. The effect of coolant blowing ratio was studied for various blowing ratios. For each blowing ratios, wall temperatures around the surface of test model were calculated. Temperature was visualized using cartesian cut-cell method to obtain traces of the injected secondary air on the test surface, so we could interpret the film effectiveness as temperature distributions.

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An Experimental Study on the Supplemental Cooling and Heating Performance Using 1 kW Thermoelectric Module for Vehicle (열전모듈을 이용한 자동차용 1 kW급 보조 냉난방 시스템의 성능에 관한 실험적 연구)

  • Lee, Dae-Woong
    • Korean Journal of Air-Conditioning and Refrigeration Engineering
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    • v.26 no.5
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    • pp.224-230
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    • 2014
  • The purpose of this paper is to investigate the performance of supplemental cooling and heating system equipped with the 1 kW thermoelectric module. The system consist of 96 thermoelectric modules, heat sink with louver fin and water cooling jacket which is attached on the hot side of the thermoelectric module. The cooling and heating performance test of the thermoelectric system is conducted with various conditions, such as intake voltage, air inlet temperature, air flow volume, water inlet temperature and water flow rate at calorimeter chamber in consideration of environmental conditions in realistic vehicle drive. The experimental results of a thermoelectric system shows that the cooling capacity and COP is 1.03 kW, and 1.0, and heating capacity and COP is 1.53 kW, and 1.5 respectively.

Influences of Cooling Conditions on the Thermally-Induced Birefringence in Injection Molding (사출성형 냉각조건이 열에 의한 복굴절에 미치는 영향)

  • Lee, H.S.;Isayev, A.I.
    • Proceedings of the Korean Society for Technology of Plasticity Conference
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    • 2007.10a
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    • pp.258-261
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    • 2007
  • Simulations of the thermally-induced residual stresses and birefringence in freely quenched plates of polycarbonate were performed by using the linear viscoelastic and photoviscoelastic constitutive equations for the mechanical and optical properties, respectively, and the first order rate equation for volume relaxation. The predictions for the birefringence showed good agreement with experimental measurements. Based on the simulation, the influences of various cooling conditions on the residual stress and birefringence in plates were investigated. The residual stress and birefringence increased with increasing initial temperature, decreasing coolant temperature and increasing heat transfer coefficient of coolants.

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Safety margin and fuel cycle period enhancements of VVER-1000 nuclear reactor using water/silver nanofluid

  • Saadati, Hassan;Hadad, Kamal;Rabiee, Ataollah
    • Nuclear Engineering and Technology
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    • v.50 no.5
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    • pp.639-647
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    • 2018
  • In this study, the effects of selecting water/silver nanofluid as both a coolant and a reactivity controller during the first operating cycle of a light water nuclear reactor are investigated. To achieve this, coupled neutronic-thermo-hydraulic analysis is employed to simulate the reactor core. A detailed VVER1000/446 reactor core is modeled in monte carlo code (MCNP), and the model is verified using the porous media approach. Results show that the maximum required level of silver nanoparticles is 1.3 Vol.% at the beginning of the cycle; this value drops to zero at the end of cycle. Due to substitution of water/boric acid with water/Ag nanofluid, reactor operation time at maximum power extends to 357.3 days, and the energy generation increases by about 27.3%. The higher negative coolant temperature coefficient of reactivity in the presence of nanofluid in comparison with the water/boric acid indicates that the reactor is inherently safer. Considering the safety margins in the presence of the nanofluid, minimum departure from nucleate boiling ratio is calculated to be 2.16 (recommendation is 1.75).

Development and verification of PWR core transient coupling calculation software

  • Li, Zhigang;An, Ping;Zhao, Wenbo;Liu, Wei;He, Tao;Lu, Wei;Li, Qing
    • Nuclear Engineering and Technology
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    • v.53 no.11
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    • pp.3653-3664
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    • 2021
  • In PWR three-dimensional transient coupling calculation software CORCA-K, the nodal Green's function method and diagonal implicit Runge Kutta method are used to solve the spatiotemporal neutron dynamic diffusion equation, and the single-phase closed channel model and one-dimensional cylindrical heat conduction transient model are used to calculate the coolant temperature and fuel temperature. The LMW, NEACRP and PWR MOX/UO2 benchmarks and FangJiaShan (FJS) nuclear power plant (NPP) transient control rod move cases are used to verify the CORCA-K. The effects of burnup, fuel effective temperature and ejection rate on the control rod ejection process of PWR are analyzed. The conclusions are as follows: (1) core relative power and fuel Doppler temperature are in good agreement with the results of benchmark and ADPRES, and the deviation between with the reference results is within 3.0% in LMW and NEACRP benchmarks; 2) the variation trend of FJS NPP core transient parameters is consistent with the results of SMART and ADPRES. And the core relative power is in better agreement with the SMART when weighting coefficient is 0.7. Compared with SMART, the maximum deviation is -5.08% in the rod ejection condition and while -5.09% in the control rod complex movement condition.

Study on the mixing performance of mixing vane grids and mixing coefficient by CFD and subchannel analysis code in a 5×5 rod bundle

  • Bin Han ;Xiaoliang Zhu;Bao-Wen Yang;Aiguo Liu;Yanyan Xi ;Lei Liu ;Shenghui Liu;Junlin Huang
    • Nuclear Engineering and Technology
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    • v.55 no.10
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    • pp.3775-3786
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    • 2023
  • Mixing Vane Grid (MVG) is one of the most important structures in fuel assembly due to its high performance in mixing the coolant and ultimately increasing Critical Heat Flux (CHF), which avoids the temperature rising suddenly of fuel rods. To evaluate the mixing performance of the MVG, a Total Diffusion Coefficient (TDC) mixing coefficient is defined in the subchannel analysis code. Conventionally, the TDC of the spacer grid is obtained from the combination of experiments and subchannel analysis. However, the processing of obtaining and determine a reasonable TDC is much challenging, it is affected by boundary conditions and MVG geometries. In is difficult to perform all the large and costing rod bundle tests. In this paper, the CFD method was applied in TDC analysis. A typical 5 × 5 MVG was simulated and validated to estimate the mixing performance of the MVG. The subchannel code was used to calculate the TDC. Firstly, the CFD method was validated from the aspect of pressure drop and lateral temperature distribution in the subchannels. Then the effect of boundary conditions including the inlet temperature, inlet velocities, heat flux ratio between hot and cold rods and the arrangement of hot and cold rods on MVG mixing and TDC were studied. The geometric effects on mixing are also carried out in this paper. The effect of vane pattern on mixing was investigated to determine which one is the best to represent the grid's mixing performance.

Effect of critical flow model in MARS-KS code on uncertainty quantification of large break Loss of coolant accident (LBLOCA)

  • Lee, Ilsuk;Oh, Deogyeon;Bang, Youngseog;Kim, Yongchan
    • Nuclear Engineering and Technology
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    • v.52 no.4
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    • pp.755-763
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    • 2020
  • The critical flow phenomenon has been studied because of its significant effect for design basis accidents in nuclear power plants. Transition points from thermal non-equilibrium to equilibrium are different according to the geometric effect on the critical flow. This study evaluates the uncertainty parameters of the critical flow model for analysis of DBA (Design Basis Accident) with the MARS-KS (Multi-dimensional Analysis for Reactor Safety-KINS Standard) code used as an independent regulatory assessment. The uncertainty of the critical flow model is represented by three parameters including the thermal non-equilibrium factor, discharge coefficient, and length to diameter (L/D) ratio, and their ranges are determined using large-scale Marviken test data. The uncertainty range of the thermal non-equilibrium factor is updated by the MCDA (Model Calibration through Data Assimilation) method. The updated uncertainty range is confirmed using an LBLOCA (Large Break Loss of Coolant Accident) experiment in the LOFT (Loss of Fluid Test) facility. The uncertainty ranges are also used to calculate an LBLOCA of the APR (Advanced Power Reactor) 1400 NPP (Nuclear Power Plants), focusing on the effect of the PCT (Peak Cladding Temperature). The results reveal that break flow is strongly dependent on the degree of the thermal non-equilibrium state in a ruptured pipe with a small L/D ratio. Moreover, this study provides the method to handle the thermal non-equilibrium factor, discharge coefficient, and length to diameter (L/D) ratio in the system code.

Artificial neural network for predicting nuclear power plant dynamic behaviors

  • El-Sefy, M.;Yosri, A.;El-Dakhakhni, W.;Nagasaki, S.;Wiebe, L.
    • Nuclear Engineering and Technology
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    • v.53 no.10
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    • pp.3275-3285
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    • 2021
  • A Nuclear Power Plant (NPP) is a complex dynamic system-of-systems with highly nonlinear behaviors. In order to control the plant operation under both normal and abnormal conditions, the different systems in NPPs (e.g., the reactor core components, primary and secondary coolant systems) are usually monitored continuously, resulting in very large amounts of data. This situation makes it possible to integrate relevant qualitative and quantitative knowledge with artificial intelligence techniques to provide faster and more accurate behavior predictions, leading to more rapid decisions, based on actual NPP operation data. Data-driven models (DDM) rely on artificial intelligence to learn autonomously based on patterns in data, and they represent alternatives to physics-based models that typically require significant computational resources and might not fully represent the actual operation conditions of an NPP. In this study, a feed-forward backpropagation artificial neural network (ANN) model was trained to simulate the interaction between the reactor core and the primary and secondary coolant systems in a pressurized water reactor. The transients used for model training included perturbations in reactivity, steam valve coefficient, reactor core inlet temperature, and steam generator inlet temperature. Uncertainties of the plant physical parameters and operating conditions were also incorporated in these transients. Eight training functions were adopted during the training stage to develop the most efficient network. The developed ANN model predictions were subsequently tested successfully considering different new transients. Overall, through prompt prediction of NPP behavior under different transients, the study aims at demonstrating the potential of artificial intelligence to empower rapid emergency response planning and risk mitigation strategies.

Frictional Characteristics of Silicon Graphite Lubricated with Water at High Pressure and High Temperature (고온 고압에서 물로 윤활되는 실리콘그라파이트 재질의 마찰 특성에 관한 연구)

  • Lee, Jae-Seon;Kim, Eun-Hyun;Park, Jin-Seok;Kim, Jong-In
    • Proceedings of the KSME Conference
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    • 2001.06a
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    • pp.151-156
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    • 2001
  • Experimental frictional and wear characteristics of silicon graphite materials is studied in this paper. Those specimens are lubricated with high temperature and highly pressurized water to simulate the same operating condition for the journal bearing and the thrust bearing on the main coolant pump bearing in the newly developing nuclear reactor named SMART(System-integrated Modular Advanced ReacTor). Operating condition of the bearings is realized by the tribometer and the autoclave. Friction coefficient and wear loss are analyzed to choose the best silicon graphite material. Pin on plate test specimens are used and coned disk springs are used to control the applied force on the specimens. Wear loss ana wear width are measured by a precision balance and a micrometer. The friction force is measured by the strain gauge which can be used under high temperature and high pressure. Three kinds of silicon graphite materials are examined and compared with each other, and each material shows similar but different results on frictional and wear characteristics.

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