• Nuclear Engineering and Technology
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• 제40권3호
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• pp.199-212
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• 2008

A small-break loss of coolant accident (SB-LOCA) test with a break size equivalent to a 3-inch cold leg break of the APR1400 was carried out as the first transient integral effect test using the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation). This was the first integral effect test to investigate the integral performance of the test facility and to verify its simulation capability for one of the design-basis accidents. Reasonably good thermal hydraulic data was obtained so that an integral performance of the fluid sub-systems was identified and control performance of the ATLAS was confirmed under real thermal hydraulic conditions. Based on the measured data, a post-test calculation was carried out using the best-estimate thermal hydraulic safety analysis code, MARS 3.1, and the similarity between the expected and actual data was investigated. On the whole, the post-test calculation reasonably predicts the major thermal hydraulic parameters measured during the SB-LOCA test. The obtained data will be used to enhance the simulation capability of the ATLAS and to improve an input model of the ATLAS for simulation of other target scenarios.

• 대한기계학회논문집
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• 제15권1호
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• pp.190-200
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• 1991

본 연구에서는 고온에 노출되는 열-점소성 거동의 해석을 위해 소성유동, 크 리프, 응력풀림(stress relaxation)등의 거동을 동시에 다룰 수 있는 통일구성방정식 모델에 대해 논하고 적절한 모델을 선정한다. 이 모델은 미소변형이론에 근거한 것 이므로 구조물의 거동을 소변형률(small strain)과 소회전(small rotation)의 범위내 로 가정하여 해석한다.선정된 모델에 대해서 시간변화율 형태의 방정식으로부터 유 한요소법을 통한 수치화와 사용된 구성방정식을 효율적으로 처리할 수 있는 수치해석 법상의 알고리듬을 제안한다. 제안된 알고리즘을 사용하여 유한요소법 전산코드를 적상하고, 작성된 코드를 이용하여 고온에서 하중을 받는 단순보와 국부적으로 심한 가열을 받는 구조물에 적용하여 고전적인 구성방정식으로 복합적인 해석이 어려웠던 열-점소성 거동을 효과적으로 해석할 수 있음을 보인다. 본 논문은 응력해석에 주안 점을 두었으므로 열해석에 관한 상세한 논의는 가급적 생략하기로 한다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.551-556
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• 1997

A present study addresses a loss-of-RHR event as an initiating event under specific low power or shutdown conditions. Two typical plant configurations, cold leg opening case with water-filled steam generators and pressurizer opening case with emptied steam generators, were evaluated using the RELAP5/ MOD3.2 code. The calculation was compared with the experiment conducted at ROSA-IV/LSTF in Japan. As a result, the code was capable of simulating the system transient behavior following the event. Especially, thermal hydraulic transport processes including non-condensable gas behavior were reasonably predicted with an appropriate time step and CPU time. However, there were some code deficiencies such as too large system mass errors and severe flow oscillations in core region.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.423-428
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• 1995

Simulation models have been developed to predict the overall behavior of the CANDU plant systems during normal operational transients. For real time simulation purpose, simplified thermal hydraulic models are applied with appropriate system control logics, which include primary heat transport system solver with its component models and secondary side system models. The secondary side models are mainly used to provide boundary conditions for primary system calculation and to accomodate plant power control logics. Also, for the effective use of simulation package, hardware oriented basic simulation network has been established with appropriate graphic display system. Through validation with typical plant power maneuvering cases using proven plant performance analysis computer code, the present simulation package shows reasonable capability in the prediction of the dynamic behavior of plant variables during operational transients of CANDU plant, which means that this simulation tool can be utilized as a basic framework for full scope simulation network through further improvements.

• Nuclear Engineering and Technology
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• 제54권9호
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• pp.3551-3566
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• 2022

The STELLA program was launched to support the PGSFR development in 2012 and for the 2nd stage, the STELLA-2 facility was designed to investigate the integral effect of safety systems including the comprehensive interaction among PHTS, IHTS and DHRS. In STELLA-2, the long-term transient behavior after accidents can be observed and the overall safety aspect can also be evaluated. In this paper, the basic design concept from engineering basis to specific design is described. The design was aimed to meet similarity criteria and requirements based on various non-dimensional numbers and the result satisfied the key features to explain the reasoning of safety evaluation. The result of this study was used to construct the facility and the experiment is on-going. In general, the final design meets the similarity criteria of the multidimensional physics inside the reactor pool. And also, for the conservation of natural circulation phenomena, the design meets the similarity requirements of geometry and thermo-dynamic behavior.

• International Journal of Concrete Structures and Materials
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• 제10권sup3호
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• pp.19-31
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• 2016

The objective of this paper is to investigate the effect of thickness and moisture on temperature distributions of reinforced concrete walls under fire conditions. Toward this goal, the first three wall specimens having different thicknesses are heated for 2 h according to ISO standard heating curve and the temperature distribution through the wall thickness is measured. Since the thermal behavior of the tested walls is influenced by thickness, as well as moisture content, three additional walls are prepared and preheated to reduce moisture content and then tested under fire exposure. The experimental results clearly show the temperatures measured close to the fire exposed surface of the thickest wall with 250 mm thickness is the highest in the temperatures measured at the same location of the thinner wall with 150 mm thickness because of the moisture clog that is formed inside the wall with 250 mm of thickness. This prevents heat being transferred to the opposite side of the heated surface. This is also confirmed by the thermal behavior of the preheated walls, showing that the temperature is well distributed in the preheated walls as compared to that in non-preheated walls. Finite element models including moisture clog zone are generated to simulate fire tests with consideration of moisture clog effect. The temperature distributions of the models predicted from the transient heat analyses are compared with experimental results and show good agreements. In addition, parametric studies are performed with various moisture contents in order to investigate effect of moisture contents on the thermal behaviors of the concrete walls.

• 한국전산유체공학회지
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• 제21권2호
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• pp.90-98
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• 2016

A numerical study is conducted on thermal performance of a large-area hot plate specially designed as a heating and cooling tool for thermal nanoimprint lithography process. The hot plate has a dimension of $240mm{\times}240mm{\times}20mm$, in which a series of cartridge heaters and cooling holes are installed. The material is stainless steel selected for enduring the high molding pressure. A numerical model based on the ANSYS Fluent is employed to predict the thermal behavior of the hot plate both in heating and cooling phases. The PID thermal control of the device is modeled by adding user defined functions. The results of numerical computation demonstrate that the use of cartridge heaters provides sufficient heat-up performance and the active liquid cooling in the cooling holes provides the required cool-down performance. However, a crucial technical issue is raised that the proposed design poses a large temperature non-uniformity in the steady heating phase and in the transient cooling phase. As a remedy, a new hot plate in which heat pipes are installed in the cooling holes is considered. The numerical results show that the installation of heat pipes could enhance the temperature uniformity both in the heating and cooling phases.

• Smart Structures and Systems
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• 제16권5호
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• pp.835-851
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• 2015

This paper presents theoretical and experimental evaluation of the structural health monitoring (SHM) capability of piezoelectric wafer active sensors (PWAS) at elevated temperatures. This is important because the technologies for structural sensing and monitoring need to account for the thermal effect and compensate for it. Permanently installed PWAS transducers have been One of the extensively employed sensor technologies for in-situ continuous SHM. In this paper, the electro-mechanical impedance spectroscopy (EMIS) method has been utilized as a dynamic descriptor of PWAS behavior and as a high frequency standing wave local modal technique. Another SHM technology utilizes PWAS as far-field transient transducers to excite and detect guided waves propagating through the structure. This paper first presents how the EMIS method is used to qualify and quantify circular PWAS resonators in an increasing temperature environment up to 230 deg C. The piezoelectric material degradation with temperature was investigated and trends of variation with temperature were deduced from experimental measurements. These effects were introduced in a wave propagation simulation software called Wave Form Revealer (WFR). The thermal effects on the substrate material were also considered. Thus, the changes in the propagating guided wave signal at various temperatures could be simulated. The paper ends with summary and conclusions followed by suggestions for further work.

• 한국전산유체공학회지
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• 제21권4호
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• pp.71-77
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• 2016

The CUPID code is a transient, three-dimensional, two-fluid, thermal-hydraulic code designed for a component-scale analysis of nuclear reactor components. The primary objective of this study is to assess the applicability of CUPID to single-phase turbulent flow analyses of $2{\times}2$ rod bundle subchannel. The bulk velocity at the inlet varies from 1.0 m/s up to 2.0 m/s which is equivalent to the fully turbulent flow with the range of Re=12,500 to 25,000. Adiabatic single-phase flow is assumed. The velocity profile at the exit region is quantitatively compared with both experimental measurement and commercial CFD tool. Three different boundary conditions are simulated and quantitatively compared each other. The calculation results of CUPID code shows a good agreement with the experimental data. It is concluded that the CUPID code has capability to reproduce the turbulent flow behavior for the $2{\times}2$ rod bundle geometry.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.72-83
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• 2022

This paper presents a numerical investigation of two-phase natural circulation flows established when external reactor vessel cooling is applied to a severe accident of the APR1400 reactor for the in-vessel retention of the core melt. The coolability limit due to external reactor vessel cooling is associated with the natural circulation flow rate around the lower head of the reactor vessel. For an elaborate prediction of the natural circulation flow rate using a thermal-hydraulic system code, MARS-KS1.5, a three-dimensional computational fluid dynamics (CFD) simulation is conducted to estimate the flow rate and pressure distribution of a liquid-state coolant at the brink of significant void generation. The CFD calculation results are used to determine the loss coefficient at major flow junctions, where substantial pressure losses are expected, in the nodalization scheme of the MARS-KS code such that the single-phase flow rate is the same as that predicted via CFD simulations. Subsequently, the MARS-KS analysis is performed for the two-phase natural circulation regime, and the transient behavior of the main thermal-hydraulic variables is investigated.