• Nuclear Engineering and Technology
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• 제53권3호
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• pp.763-775
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• 2021

The thermal hydraulic performances of Steam Generator (SG) under both steady and transient operation conditions are of great importance for the safety and economy in nuclear power plants. In this paper, based on our self-developed SG thermal hydraulic analysis code STAF (Steam-generator Thermalhydraulic Analysis code based on Fluent), an improved new version STAF-CT (fully Coupling and Transient) is developed and introduced. Compared with original STAF, the new version code STAF-CT has two main functional improvements including "Transient" and "Fully Three Dimensional Coupling" features. In STAF-CT, a three dimensional energy transferring module is established which can achieve energy exchange computing function at the corresponding position between two sides of SG. The STAF-CT is validated against the international benchmark experiment data and the results show great agreement. Then the U-shaped SG in AP1000 nuclear power plant is modeled and simulated using STAF-CT. The results show that three dimensional flow fields in the primary side make significant effect on the energy source distribution between two sides. The development of code STAF-CT in this paper can provide an effective method for further SG high fidelity research in the nuclear reactor system.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.3990-4002
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• 2021

CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.310-323
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• 2023

The k-eigenvalue neutronics/thermal-hydraulics coupling calculation is a key issue for reactor design and analysis. Jacobian-free Newton-Krylov (JFNK) method, featured with super-linear convergence rate and high efficiency, has been attracting more and more attention to solve the multi-physics coupling problem. However, it may converge to the high-order eigenmode because of the multiple solutions nature of the k-eigenvalue form of multi-physics coupling issue. Based on our previous work, a modified JFNK with a line search method is proposed in this work, which can find the fundamental eigenmode together with thermal-hydraulics feedback in a wide range of initial values. In detail, the existing modified JFNK method is combined with the line search strategy, so that the intermediate iterative solution can avoid a sudden divergence and be adjusted into a convergence basin smoothly. Two simplified 2-D homogeneous reactor models, a PWR model, and an HTR model, are utilized to evaluate the performance of the newly proposed JFNK method. The results show that the performance of this proposed JFNK is more robust than the existing JFNK-based methods.

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 2004년도 춘계 학술발표대회 개요집
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• pp.66-68
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• 2004

Induction heating is a process that is accompanied with magnetic and thermal situation. When the high-frequency current flows in the coil, induced eddy current generates heat to conductor. To simulate an induction heating process, the finite element analysis program was developed. A coupling method between the magnetic and thermal routines was developed. In the process of magnetic analysis and thermal analysis, magnetic material properties and thermal material properties depending on temperature are taken into consideration. In this paper, to predict the angular deformation, temperature difference and the shape of heat affected zone were discussed. Also appropriate coil shape for maximum angular deformation were proposed.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.641-646
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• 1998

In order to enhance the dynamic and interactive simulation capability of a system thermal hydraulic code for nuclear power plant, applicability of flow network models in SINDA/FLUIN $T^{™}$ has been tested by modeling feedwater system and coupling to DSNP which is one of a system thermal hydraulic simulation code for a pressurized heavy water reactor. The feedwater system is selected since it is one of the most important balance of plant systems with a potential to greatly affect the behavior of nuclear steam supply system. The flow network model of this feedwater system consists of condenser, condensate pumps, low and high pressure heaters, deaerator, feedwater pumps, and control valves. This complicated flow network is modeled and coupled to DSNP and it is tested for several normal and abnormal transient conditions such turbine load maneuvering, turbine trip, and loss of class IV power. The results show reasonable behavior of the coupled code and also gives a good dynamic and interactive simulation capabilities for the several mild transient conditions. It has been found that coupling system thermal hydraulic code with a flow network code is a proper way of upgrading simulation capability of DSNP to mature nuclear plant analyzer (NPA).

• Elastomers and Composites
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• 제46권3호
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• pp.237-244
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• 2011

비보강, 카본블랙, 카본블랙과 실란커플링제, 실리카, 실리카와 실란커플링제 등 5가지의 다른 충전 시스템을 갖는 NR 복합체를 열노화시켰으며, 가속 열노화에 의한 가교밀도 변화를 연구하였다. 가교밀도는 충전 시스템과 관계없이 노화 시간이 경과함에 따라 증가하였다. 노화 온도가 높아질수록 가교밀도는 눈에 띄게 변하였다. 카본블랙으 로 보강된 NR 복합체의 경우, 실란커플링제는 가교밀도를 증가시켰다. 하지만 실리카로 보강된 NR 복합체의 경우에는 $60^{\circ}C$와 $70^{\circ}C$ 노화에서는 가교밀도 증가분이 감소하였으며, $80^{\circ}C$와 $90^{\circ}C$에서는 거의 영향을 주지 않았다. 카본블랙으로 보강된 NR 복합체의 가교밀도 변화에 대한 활성화 에너지는 로그 함수 형태로 지속적으로 증가한 반면, 실리카로 보강된 NR 복합체의 경우에는 노화 시간 30-150일 구간에서는 거의 변화가 없었다. 비보강 시험편의 활성화 에너지 는 노화 시간에 따라 지수 함수 형태로 증가하였다. 실험 결과는 실란커플링제로부터의 황 제공, 실란커플링제에 의한 충전제의 표면 개질, 실리카 표면에서의 가교제 잔류물의 흡착, 그리고 흡착된 가교제 잔류물의 방출 등으로 설명하였다.

• Tribology and Lubricants
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• 제25권6호
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• pp.421-426
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• 2009

The brake system is important part of automobile safety system. The disc brake system is divided two parts: the rotating axisymmetrical disc and the stationary pads. During braking, the kinetic energy and potential energy of moving vehicle were converted into the thermal energy through frictional heat between the brake disc and the pads. The frictional heat, which is generated on the interface of the disc and pads, can cause high temperature during the braking process. The object of present work is to determine temperature and thermal stress, to compare to simulation results and experimental results in the disc by partial 3D model of ventilated disc brake with appropriate boundary conditions. In the simulation process, the mechanical loads were applied to the thermo-mechanical coupling analysis in order to simulate the process of heat produced by friction.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.2828-2839
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• 2022

This study developed a safety assessment tool for geological disposal systems called APro, a systemically integrated modeling system based on modularizing and coupling the processes which need to be considered in a geological disposal system. Thermal, hydraulic, chemical, canister failure, radionuclide release and transport processes were considered in the current version of APro. Each of the unit processes in APro consists of a single Default Module, and several Alternative Modules which can increase the flexibility of the model. As an initial stage of developing the modularization concept and modeling interface, the Default Modules of each unit process were described, with one Alternative Module of chemical process. The computation part of APro is mainly a MATLAB workspace controlling COMSOL and PHREEQC, which are coupled by an operator splitting scheme. The APro model domain is a stylized geological disposal system employing the Swedish disposal concept (KBS-3 type), but the repository layout can be freely adjusted. In order to show the applicability of APro to the total system performance assessment of geological disposal system, some sample simulations were conducted. From the results, it was confirmed that coupling of the thermal and hydraulic processes and coupling of the canister failure and the radionuclide release processes were well reflected in APro. In addition, the technical connectivity between COMSOL and PHREEQC was also confirmed.

• 센서학회지
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• 제26권5호
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• pp.338-341
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• 2017

We have investigated a temperature sensor on a thermal expansion effect of a fused optical fiber coupler. Both side of the fused tapered region of the coupler were fixed on a metal support to induce the high thermal expansion effect. The sensor showed that the peak coupling wavelengths were shifted to shorted wavelength region with increased of environmental temperature. The sensitivity of the sensor was $0.12nm/^{\circ}C$.

• 호남수학학술지
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• 제38권3호
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• pp.435-454
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• 2016

This paper investigates the effect of dual-phase-lags on a thermoviscoelastic orthotropic solid with a cylindrical cavity. The cylindrical cavity is subjected to a thermal shock varying heat and its material is taken to be of Kelvin-Voigt type. The phase-lag thermoelastic model, Lord and Shulman's model and the coupled thermoelasticity model are employed to study the thermomechanical coupling, thermal and mechanical relaxation (viscous) effects. Numerical solutions for temperature, displacement and thermal stresses are obtained by using the method of Laplace transforms. Numerical results are plotted to illustrate the effect phase-lags, viscoelasticity, and the variability thermal conductivity parameter on the studied fields. The variations of all field quantities in the context of dual-phase-lags and coupled thermoelasticity models follow similar trends while the Lord and Shulman's model may be different. The influence of viscosity parameter and variability of thermal conductivity is very pronounced on temperature and thermal stresses of the thermoviscoelastic solids.