• Nuclear Engineering and Technology
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• 제49권5호
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• pp.968-978
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• 2017

Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal-hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

• Tribology and Lubricants
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• 제25권2호
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• pp.102-107
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• 2009

The thermal deformation volume of motorcycle break disk was studied using a disk-on-pad type friction tester. Thermal deformation volume of motorcycle break disk have an effect on the frictional factor such as applied load, sliding speed, sliding distance and number of ventilated disk hole. However, it is difficult to know the mutual relation of these factors on thermal deformation volume. In this study, the thermal deformation volume with ANSYS workbench are obtained by application of temperature from mechanical test. From this study, the result was shown that the motorcycle break disk with ventilated hole 3 have the most excellent thermal deformation characteristics. The regression equation with frictional factors which have a trust rate of 95% for prediction of thermal deformation volume of motorcycle break disk was composed.

• Nuclear Engineering and Technology
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• 제41권5호
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• pp.655-676
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• 2009

The first-ever integral effect test for simulating a guillotine break of a DVI (Direct Vessel Injection) line of the APR1400 was carried out with the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) from the same prototypic pressure and temperature conditions as those of the APR1400. The major thermal hydraulic behaviors during a DVI line break accident were identified and investigated experimentally. A method for estimating the break flow based on a balance between the change in RCS inventory and the injection flow is proposed to overcome a direct break low measurement deficiency. A post-test calculation was performed with a best-estimate safety analysis code MARS 3.1 to examine its prediction capability and to identify any code deficiencies for the thermal hydraulic phenomena occurring during the DVI line break accidents. On the whole, the prediction of the MARS code shows a good agreement with the measured data. However, the code predicted a higher core level than did the data just before a loop seal clearing occurs, leading to no increase in the peak cladding temperature. The code also produced a more rapid decrease in the downcomer water level than was predicted by the data. These observable disagreements are thought to be caused by uncertainties in predicting countercurrent flow or condensation phenomena in a downcomer region. The present integral effect test data will be used to support the present conservative safety analysis methodology and to develop a new best-estimate safety analysis methodology for DVI line break accidents of the APR1400.

• 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.

• Tribology and Lubricants
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• 제24권6호
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• pp.308-314
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• 2008

The thermal stress have an effect on the frictional factor such as applied load, sliding speed, sliding distance and number of ventilated disk hole. However, it is difficult to know the mutual relation of these factors on thermal stress of motorcycle break disk. For this, temperature of motorcycle break disk is measured using a disk-on-pad type friction tester with full factorial design containing above 4 elements. and the thermal stress analysis of it was carried out using with ANSYS workbench. From this study, the result was shown that the regression equation which have a trust rate of 95% for thermal stress presumption of motorcycle break disk with frictional factor was composed. It is possible to apply for another automobile parts.

• Nuclear Engineering and Technology
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• 제41권6호
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• pp.775-784
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• 2009

This paper summarizes the tests performed in the ATLAS facility during its first two years of operation (2007${\sim}$2008). Two categories of tests have been performed successfully: (a) the reflood phase of the large-break loss-of-coolant accidents in a cold leg, and (b) the breaks in one of four direct vessel injection lines. Those tests contributed to understanding the unique thermal-hydraulic behavior, resolving the safety-related concerns and providing an evaluation of the safety analysis codes and methodology for the advanced pressurized water reactor, APR1400. Several important and interesting phenomena have been observed during the tests. In most cases, the ATLAS shows reasonable accident characteristics and conservative results compared with those predicted by one-dimensional safety analysis codes. A wide variety of small-break LOCA tests will be performed in 2009.

• 전자공학회논문지
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• 제53권10호
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• pp.109-115
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• 2016

이 논문에서는 레이저를 통한 진동판 스캔과 열화상 카메라를 사용한 진동판 촬영, 두 가지 방법을 비교하며 슬림 스피커의 분할 진동을 검출하는 방법에 대해 살펴본다. 슬림 스피커는 평판형의 구조적인 특성상 분할진동이 두드러지게 나타나고, 설치되는 공간이 좁아 무빙 코일에서 발생하는 열의 냉각이 제한적이다. 이런 특성으로 인해 슬림 스피커에서 분할진동이 제품의 품질에 큰 영향을 미친다. 본 연구에서는 진동판에서 일어나는 분할진동의 영향과 무빙 코일에 의한 진동판의 열전달 관계를 비교 탐색한다. 비교를 위한 실험은 분할진동 모드의 측정과 진동판의 열 변화 측정을 진동판 스캔과 열화상 카메라 촬영의 2단계의 실험으로 진행한다. 동일 주파수에서 발생하는 분할진동 모드와 열전달 형태를 비교하여 서로 간에 어떤 영향을 미치고 있는 지 파악할 수 있다. 그리고 이를 통해 발견한 연관성을 통해 쉽게 촬영할 수 있는 열화상만으로도 슬림 스피커가 가지는 분할진동의 형태와 경향성을 빠르게 파악하여 최적 설계에 도움이 되는 자료로 사용할 수 있을 것이다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.635-640
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• 1998

Thermal fluid mixing behavior during a postulated medium-size hot leg break loss of coolant accident is analyzed for the international comparative assessment study on pressurized thermal shock (PTS-ICAS) proposed by OECD-NEA. The applicability of RELAP5 code to analyze thermal fluid mixing behavior is evaluated through a simple modeling relevant to the problem constraints. Based on the calculation result, the onset of Thermal stratification is investigated using Theofanous's empirical correlation. Sensitivity calculations using a fine node model and crossflow model are also performed to evaluate the modeling capability on multi-dimensional characteristics related to thermal fluid mixing.

• Nuclear Engineering and Technology
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• 제47권6호
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• pp.712-718
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• 2015

Cast duplex stainless steels are susceptible to thermal aging during long-term service at temperatures ranging from $280^{\circ}C$ to $450^{\circ}C$. To analyze the effect of thermal aging on leak-before-break (LBB) behavior, three-dimensional finite element analysis models were built for circumferentially cracked pipes. Based on the elasticeplastic fracture mechanics theory, the detectable leakage crack length calculation and J-integral stability assessment diagram approach were carried out under different bending moments. The LBB curves and LBB assessment diagrams for unaged and thermally aged pipes were constructed. The results show that the detectable leakage crack length for thermally aged pipes increases with increasing bending moments, whereas the critical crack length decreases. The ligament instability line and critical crack length line for thermally aged pipes move downward and to the left, respectively, and unsafe LBB assessment results will be produced if thermal aging is not considered. If the applied bending moment is increased, the degree of safety decreases in the LBB assessment.

• Nuclear Engineering and Technology
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• 제39권4호
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• pp.287-298
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• 2007

As part of the $AP1000^{TM}$ pressurized water reactor design certification program, a series of integral systems tests of the nuclear steam supply system was performed at the APEX-1000 test facility at Oregon State University. These tests provided data necessary to validate Westinghouse safety analysis computer codes for AP1000 applications. In addition, the tests provided the opportunity to investigate the thermal-hydraulic phenomena expected to be important in AP1000 small-break loss of coolant accidents (SBLOCAs). The APEX-1000 facility is a 1/4-scale pressure and 1/4-scale height simulation of the AP1000 nuclear steam supply system and passive safety features. A series of eleven tests was performed in the APEX-1000 facility as part of a U.S. Department of Energy contract. In all, four SBLOCA tests representing a spectrum of break sizes and locations were simulated along with tests to study specific phenomena of interest. The focus of this paper is the SBLOCA tests. The key thermal-hydraulic phenomena simulated in the APEX-1000 tests, and the performance and interactions of the passive safety-related systems that can be investigated through the APEX-1000 facility, are emphasized. The APEX-1000 tests demonstrate that the AP1000 passive safety-related systems successfully combine to provide a continuous removal of core decay heat and the reactor core remains covered with considerable margin for all small-break LOCA events.