• 한국기계가공학회지
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• 제17권4호
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• pp.150-159
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• 2018

The present research deals with a finite element analysis and fatigue evaluation of a steam separator of a high-pressure evaporator for the Heat Recovery Steam Generator (HRSG). The fatigue during the expected life of the HRSG was evaluated according to the ASME Boiler and Pressure Vessel Code Section VIII Division 2 (ASME Code). First, based on the eight transient operating conditions prescribed for the HRSG, temperature distribution of the steam separator was analyzed by a transient thermal analysis. Results of the thermal analysis were used as a thermal load for the structural analysis and used to determine the mean cycle temperature. Next, a structural analysis for the transient conditions was carried out with the thermal load, steam pressure, and nozzle load. The maximum stress location was found to be the riser nozzle bore, and hence fatigue was evaluated at that location, as per ASME Code. As a result, the cumulative usage factor was calculated as 0.00072 (much less than 1). In conclusion, the steam separator was found to be safe from fatigue failure during the expected life.

• 한국자동차공학회논문집
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• 제19권6호
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• pp.126-132
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• 2011

Fatigue cracks of the turbocharger are often observed for high performance engines under thermal shock tests. Maximum exhaust gas temperature of recently developed gasoline engines could reach approximately $950^{\circ}C$. It's very important to estimate transient temperature histories during thermal shock cycles to predict the stress and the fatigue life of the turbocharger. With these temperature profiles, temperature-dependent material properties and boundary conditions, we could identify critical locations by the application of finite element simulation technologies. In this paper, we applied the reliable analysis approach to the actual turbocharger to predict the weak locations due to the repetitions of plastic strains and compared the results with the crack locations under physical engine test.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회B
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• pp.2346-2351
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• 2008

This study deals with heat pipes inserted into the metal hydride(MH) reactor to increase the effective thermal conductivity of the system and thus to enhance the thermal control characteristics. A numerical analysis was conducted to predict the effect of inserted heat pipes on the heat transfer characteristics of MH, which inherently has extremely low thermal conductivity. The numerical model was a cylindrical container of O.D. 76.3 mm and length 1 m, which is partially filled with about 60% of MH material. The heat pipe was made of copper-water combination, which is suitable for operation temperature range between $10^{\circ}C$ and $80^{\circ}C$. Both inner -and outer- heat pipes were considered in the model. Less than two hours of transient time is of concern when decreasing or increasing the temperature for absorption and discharge of hydrogen gas. FLUENT, a commercial software, was employed to predict the transient as well as steady-state temperature distribution of the MH reactor system. The numerical results were compared and analyzed from the view point of temperature uniformity and transient time up to the specified maximum or minimum temperatures.

• Nuclear Engineering and Technology
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• 제45권4호
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• pp.439-458
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• 2013

REX-10 is a fully-passive small modular reactor in which the coolant flow is driven by natural circulation, the RCS is pressurized by a steam-gas pressurizer, and the decay heat is removed by the PRHRS. To confirm design decisions and analyze the transient responses of an integral PWR such as REX-10, a thermal-hydraulic system code named TAPINS (Thermal-hydraulic Analysis Program for INtegral reactor System) is developed in this study. Based on a one-dimensional four-equation drift-flux model, TAPINS incorporates mathematical models for the core, the helical-coil steam generator, and the steam-gas pressurizer. The system of difference equations derived from the semi-implicit finite-difference scheme is numerically solved by the Newton Block Gauss Seidel (NBGS) method. TAPINS is characterized by applicability to transients with non-equilibrium effects, better prediction of the transient behavior of a pressurizer containing non-condensable gas, and code assessment by using the experimental data from the autonomous integral effect tests in the RTF (REX-10 Test Facility). Details on the hydrodynamic models as well as a part of validation results that reveal the features of TAPINS are presented in this paper.

• 대한기계학회논문집B
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• 제35권2호
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• pp.113-120
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• 2011

본 논문은 나노유체의 열확산율을 측정하는 센서와 주변회로 그리고 데이터의 처리방법을 제시한 것이다. 기존 비정상열선법을 이용하면 이론상 유체의 열전도율과 열확산율을 동시에 측정할 수 있으나 열전도율과 비교하여 열확산율은 많은 오차가 발생한다. 본 연구에서 제시한 방법은 측정변수가 단순하고 복잡한 센서의 교정과정을 생략할 수 있는 실용적 측면의 장점이 있다. 먼저 열확산율이 잘 알려진 유체들에 대한 검증실험을 실시하였고 나노유체의 열확산율을 측정하여 기본유체와 비교하는 과정을 예시적으로 설명하였다. 본 연구는 기존 열전도율측정에 한정되어 왔던 나노유체 연구의 범위를 열확산율 또는 비열의 개념으로 확장하였다는 관점에서 중요성을 갖는다.

• 항공우주시스템공학회지
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• 제10권2호
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• pp.1-6
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• 2016

The thermal analysis of electronic equipment is required to predict the reliability of electronic equipment being loaded on a satellite. The transient heat transfer of electronic equipment that was developed recently has been generated using a large-scale integration circuit. If there is a transient heat transfer between EEE(Electric, Electronic and Electro mechanical) parts, it may lead to failure the satellite mission. In this study, we performed the thermal design and analysis for reliability of APD(Antenna Pointing Driver) electronics for activation of a spaceborne X-band 2-axis antenna. The EEE parts were designed using a thermal mathematical model without the thermal mitigation element. In addition, thermal analysis was performed based on the worst case for verifying the reliability of EEE parts. For the thermal analysis results, the thermal stability of electronic equipment has been demonstrated by satisfying the de-rating junction temperature.

• 한국자동차공학회논문집
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• 제18권1호
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• pp.131-138
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• 2010

Because real flow of engine exhaust is very hot and highly transient, it may cause thermal and inertial loads on catalyzed filters in DPF. Transient and detailed flow and thermal simulations are necessary in this field. To assess the importance of time dependent phenomena, typical cone-type configuration such as an underbody DPF is selected for steady and transient analysis. User defined functions of FLUENT by sinusoidal inlet velocities are written and integrated with main solver for realistic simulation. Also, 4-cylinder and 6-cylinder engines for 3,000 L class are considered for the dynamic exhaust effect of engine type. Key parameters to understanding of catalyst performance and durability issues such as flow uniformity index and peak velocity are investigated. Also, pressure drop for engine power are considered. From the simulation results for three different cases, proper approach is recommended.

• Nuclear Engineering and Technology
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• 제52권6호
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• pp.1110-1119
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• 2020

The Floating Absorber for Safety at Transient (FAST) is a safety device used in the innovative Sodium-cooled Fast Reactor (iSFR). The FAST insert negative reactivity under transient or accident conditions. However, behavior of the FAST is still unclear under transient conditions. Therefore, the existing Floating Absorber for Safety at Transient Analysis Code (FASTAC) is improved to analyze the FAST movement by considering the reactivity and temperature distribution within the reactor core. The current FAST system is simulated under a single control rod withdrawal accident condition. In this investigation, the reactor thermal power does not return to its initial thermal power even if the FAST inserts negative reactivity. Only a 9 K of coolant temperature margin, in the hottest fuel assembly at EOL, can lead to unnecessary insertion of the negative reactivity. On the other hand, the FASTs cannot contribute to controlling the reactivity when normalized radial power is less than 0.889 at BOL and 0.972 at EOL. These simulation results suggest that the current FAST design needs to be optimized depending on its installed location. Meanwhile, the FAST system keeps the fuel, cladding and coolant temperatures below their limit temperatures with given conditions.

• 한국전산유체공학회지
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• 제17권4호
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• pp.41-48
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• 2012

A numerical analysis of thermal stratification in the upper plenum of the MONJU fast breeder reactor was performed. Calculations were performed for a 1/6 simplified model of the MONJU reactor using the commercial code, CFX-13. To better resolve the geometrically complex upper core structure of the MONJU reactor, the porous media approach was adopted for the simulation. First, a steady state solution was obtained and the transient solutions were then obtained for the turbine trip test conducted in December 1995. The time dependent inlet conditions for the mass flow rate and temperature were provided by JAEA. Good agreement with the experimental data was observed for steady state solution. The numerical solution of the transient analysis shows the formation of thermal stratification within the upper plenum of the reactor vessel during the turbine trip test. The temporal variations of temperature were predicted accurately by the present method in the initial rapid coastdown period (~300 seconds). However, transient numerical solutions show a faster thermal mixing than that observed in the experiment after the initial coastdown period. A nearly homogenization of the temperature field in the upper plenum is predicted after about 900 seconds, which is a much shorter-term thermal stratification than the experimental data indicates. This discrepancy is due to the shortcoming of the turbulence models available in the CFX-13 code for a natural convection flow with thermal stratification.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2343-2351
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• 2024

Accurately predicting the thermal hydraulic parameters of a transient reactor core under different working conditions is the first step toward reactor safety. Mass flow rate and temperature are important parameters of core thermal hydraulics, which have often been modeled as time series prediction problems. This study aims to achieve accurate and continuous prediction of core thermal hydraulic parameters under instantaneous conditions, as well as test the feasibility of a newly constructed gated recurrent unit (GRU) model based on the soft attention mechanism for core parameter predictions. Herein, the China Experimental Fast Reactor (CEFR) is used as the research object, and CEFR 1/2 core was taken as subject to carry out continuous predictive analysis of thermal parameters under transient conditions., while the subchannel analysis code named SUBCHANFLOW is used to generate the time series of core thermal-hydraulic parameters. The GRU model is used to predict the mass flow and temperature time series of the core. The results show that compared to the adaptive radial basis function neural network, the GRU network model produces better prediction results. The average relative error for temperature is less than 0.5 % when the step size is 3, and the prediction effect is better within 15 s. The average relative error of mass flow rate is less than 5 % when the step size is 10, and the prediction effect is better in the subsequent 12 s. The GRU model not only shows a higher prediction accuracy, but also captures the trends of the dynamic time series, which is useful for maintaining reactor safety and preventing nuclear power plant accidents. Furthermore, it can provide long-term continuous predictions under transient reactor conditions, which is useful for engineering applications and improving reactor safety.