• Nuclear Engineering and Technology
• /
• 제53권9호
• /
• pp.3112-3121
• /
• 2021

One of the major factors affecting the life span of a Reactor Pressure Vessel (RPV) is the Pressurised Thermal Shock (PTS). PTS is a thermo-mechanical load on the RPV wall due to steep temperature gradients and structural load created by internal pressure of the fluid within the RPV. Safe operating life of a nuclear power plant is ensured by carrying out fracture analysis of the RPV against thermal shock. Carrying out fracture tests on RPV/large scale components is not always feasible. Hence, studies on laboratory level specimens are necessary to validate and supplement the prototype results. This paper aims to study the fracture behaviour of standard Compact Tension [C(T)] specimens, made of RPV steel 20MnMoNi55, subjected to thermal shock through experimental and numerical investigations. Fracture tests have been carried out on the C(T) specimens subjected to thermal transient load and tensile load to quantify the effect of thermal shock. Crack resistance curves are obtained from the fracture tests as per ASTM E1820 and compared with those obtained numerically using XFEM and a good agreement was found. A quantitative study on the crack tip plastic zone, computed using cohesive segment approach, from the numerical analyses justified the experimental crack initiation toughness.

• 한국항해항만학회:학술대회논문집
• /
• 한국항해항만학회 2006년도 추계학술대회 논문집(제1권)
• /
• pp.367-372
• /
• 2006

선형모터의 설계에 있어서 열적 거동의 파악은 모터의 연속 추력 및 열변형과 관련하여 중요한 고려요소 중의 하나이다. 본 연구에서는 항만자동화를 위한 차세대 컨테이너의 수평이송시스템인 LMTT(Linear Motor-based Transfer Technology)용 선형모터의 설계를 위해 고정자의 코일에서 발생되는 열에 의한 이동체의 온도분포를 분석하여 고정자 모듈의 열해석을 고려한 이동체의 구조설계에 관한 연구를 수행하였다. 먼저, 선형모터의 주요부품의 치수를 설계하고, 다음으로 고정자 모듈의 발열을 고려한 이동체의 온도분포를 분석하여 열-구조 연성해석을 수행하였다.

• 대한전기학회:학술대회논문집
• /
• 대한전기학회 2009년도 제40회 하계학술대회
• /
• pp.1613_1614
• /
• 2009

In this paper, thermal movement of high power LED on the MCPCB is analyzed with structural function and CFdesign V10 program. thermal resistance is decreased as 10.1 [$^{\circ}C$/W] in MCPCB from 12.2 [$^{\circ}C$/W] in LED package. Junction temperature which is calculated with thermal computational analysis program shows 85.113 [$^{\circ}C$] and almost same to measured data.

• 한국정밀공학회:학술대회논문집
• /
• 한국정밀공학회 2002년도 추계학술대회 논문집
• /
• pp.980-984
• /
• 2002

This paper has studied thermal characteristics of machine tool to develope high speed spindle and optimum design condidering the thermal deformation. Comparing the test data of temperature measurement and structural analysis data using FEM, we verified the test validity and predicted thermal deformation, influence of spindle generation of heat, and established cooling system to prevent the thermal deformation. 1) The temperature rise of spindle system depends on increasing number of rotation and shows sudden doubling increment of number of rotation over 7,000rpm. 2) Oil jacket cooling can be effective cooling method below 8,000rpm but, over 8,000rpm, it shows the decrement of cooling effect. 3) Comparing FEM analysis results and revolution test results, we can confirm approximate temperature change consequently, it is possible to simulate temperature rise and thermal distribution on the inside of spindle system. 4) We can confirm that simulated approach by FEM analysis can be effective method in thermal-appropriate design.

• 한국항공우주학회지
• /
• 제43권9호
• /
• pp.805-813
• /
• 2015

본 논문은 한국항공우주연구원에서 개발 중인 위성종합설계 소프트웨어의 부분들 중 열 해석 및 구조해석 격자 사이의 온도 해석 결과를 변환하는 알고리즘에 대한 전반적인 연구 내용을 기술한다. 일반적으로 인공위성은 궤도상의 다양한 우주환경에 따라서 온도가 시간에 따라 변하며 이에 따른 열변형을 고려한 구조적 안정성을 설계 단계에서 고려해야 한다. 본문에는 유한요소 열-구조 연계해석을 위한 이종 격자 간 온도 정보를 보간하는 방법을 제시하였으며 이 방법을 통해 온도 변환이 잘 수행됨을 확인할 수 있다.

• Nuclear Engineering and Technology
• /
• 제54권10호
• /
• pp.3928-3942
• /
• 2022

Thermal striping is a complex thermal-hydraulic phenomenon caused by fluid temperature fluctuations that can also cause high-cycle thermal fatigue to the structural wall of sodium-cooled fast reactors (SFRs). Numerical simulations using large-eddy simulation (LES) were performed to predict and evaluate the characteristics of the temperature fluctuations related to thermal striping in the upper internal structure (UIS) of the prototype generation-IV sodium-cooled fast reactor (PGSFR). Specific monitoring points were established for the fluid region near the control rod driving mechanism (CRDM) guide tubes, CRDM guide tube walls, and UIS support plates, and the normalized mean and fluctuating temperatures were investigated at these points. It was found that the location of the maximum amplitude of the temperature fluctuations in the UIS was the lowest end of the inner wall of the CRDM guide tube, and the maximum value of the normalized fluctuating temperatures was 17.2%. The frequency of the maximum temperature fluctuation on the CRDM guide tube walls, which is an important factor in thermal striping, was also analyzed using the fast Fourier transform analysis. These results can be used for the structural integrity evaluation of the UIS in SFR.

• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2008년도 추계학술대회A
• /
• pp.554-559
• /
• 2008

Most of vehicle manufacturers have applied exhaust gas recirculation (EGR) system to the development of diesel engines in order to obtain the high thermal efficiency without $NO_X$ and Particulate Matter (PM) emitted from the engine. EGR system, which reflow a cooled exhaust gas from vehicles burning diesel as fuel to a combustion chamber of engine, has been used to solve this problem. In order to confirm the safety of the EGR system, finite element analysis was carried out. The safety of EGR system against temperature variation in the shell and tubes was evaluated through the thermal and structural analysis, and the modal analysis using ANSYS was also performed. Finally, the performance of EGR system was verified through the experiment and numerical simulation using effectiveness-NTU method. Program for the estimation of the heat exchange efficiency of the EGR system with regard to the dimpled tube shape was developed.

• Structural Engineering and Mechanics
• /
• 제16권6호
• /
• pp.695-712
• /
• 2003

The non-linear structural analysis of reinforced concrete beams in fire consists of three separate steps: (i) The estimation of the rise of surrounding air temperature due to fire; (ii) the determination of the distribution of the temperature within the beam during fire; (iii) the evaluation of the mechanical response due to simultaneous time-dependent thermal and mechanical loads. Steps (ii) and (iii) are dealt with in the present paper. We present a two-step computational procedure where a 2D transient thermal analysis over the cross-sections of beams are made first, followed by mechanical analysis of the structure. Fundamental to the accuracy of the mechanical analysis is a new planar beam finite element. The effects of plasticity in concrete, and plasticity and viscous creep in steel are taken into consideration. The properties of concrete and steel along with the values of their thermal and mechanical parameters are taken according to the European standard ENV 1992-1-2 (1995). The comparison of our numerical and full-scale experimental results shows that the proposed mechanical and 2D thermal computational procedure is capable to describe the actual response of reinforced concrete beam structures to fire.

• Structural Engineering and Mechanics
• /
• 제37권3호
• /
• pp.257-265
• /
• 2011

Plant-specific analyses of 5 types of domestic reactors in Korea are performed to assure the structural integrity of the reactor pressure vessel (RPV) during transients which are expected to initiate pressurized thermal shock (PTS) events. The failure probability of the RPV due to PTS is obtained by performing probabilistic fracture mechanics analysis. The through-wall cracking frequency is calculated and compared to the acceptance criterion. Considering the fluence at the end of life expected by surveillance test, the sufficient safety margin is expected for the structural integrity of all reactor pressure vessels except for the oldest one during the pressurized thermal shock events. If the flaw with aspect ratio of 1/12 is considered to eliminate the conservatism, the acceptance criteria is not exceeded for all plants until the fluence level of $8{\times}10^{19}\;n/cm^2$, generating sufficient margin beyond the design life.

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