• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1412-1420
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• 2018

An experiment was performed using the large-scale test facility (LSTF), which simulated a 1% vessel upper head small-break loss-of-coolant accident with an accident management (AM) measure under an assumption of total-failure of high-pressure injection (HPI) system in a pressurized water reactor (PWR). In the LSTF test, liquid level in the upper head affected break flow rate. Coolant was manually injected from the HPI system into cold legs as the AM measure when the maximum core exit temperature reached 623 K. The cladding surface temperature largely increased due to late and slow response of the core exit thermocouples. The AM measure was confirmed to be effective for the core cooling. The RELAP5/MOD3.3 code indicated insufficient prediction of primary coolant distribution. The author conducted uncertainty analysis for the LSTF test employing created phenomena identification and ranking table for each component. The author clarified that peak cladding temperature was largely dependent on the combination of multiple uncertain parameters within the defined uncertain ranges.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.928-940
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• 2017

An experiment using the $Prim{\ddot{a}}rkreisl{\ddot{a}}ufe$ Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg smallbreak loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1537-1546
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• 2017

Many thermal-hydraulic tests have been conducted at the Korea Atomic Energy Research Institute for verification of the SMART (System-integrated Modular Advanced ReacTor) design, the standard design approval of which was issued by the Korean regulatory body. In this paper, the contributions of these tests to the standard design approval of SMART are discussed. First, an integral effect test facility named VISTA-ITL (Experimental Verification by Integral Simulation of Transients and Accidents-Integral Test Loop) has been utilized to assess the TASS/SMR-S (Transient and Set-point Simulation/Small and Medium) safety analysis code and confirm its conservatism, to support standard design approval, and to construct a database for the SMART design optimization. In addition, many separate effect tests have been performed. The reactor internal flow test has been conducted using the SCOP (SMART COre flow distribution and Pressure drop test) facility to evaluate the reactor internal flow and pressure distributions. An ECC (Emergency Core Coolant) performance test has been carried out using the SWAT (SMART ECC Water Asymmetric Two-phase choking test) facility to evaluate the safety injection performance and to validate the thermal-hydraulic model used in the safety analysis code. The Freon CHF (Critical Heat Flux) test has been performed using the FTHEL (Freon Thermal Hydraulic Experimental Loop) facility to construct a database from the $5{\times}5$ rod bundle Freon CHF tests and to evaluate the DNBR (Departure from Nucleate Boiling Ratio) model in the safety analysis and core design codes. These test results were used for standard design approval of SMART to verify its design bases, design tools, and analysis methodology.

• 한국화재소방학회논문지
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• 제23권3호
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• pp.11-16
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• 2009

본 논문은 전기설비에서 사고 발생 징후를 검출하기 위한 방법을 기술하고 있다. 실험을 위해 전기설비의 부하는 백열전구로 구성하였고, 사고는 전기설비의 일부에 KS C IEC(Korea Standard C International Electrostatic Commission) 60112에 따른 내트래킹 시험장치를 연결하여 모의되었다. 모의사고는 정상상태에서 방전진전과 탄화형성 과정을 거쳐 트래킹파괴에 이르게 된다. 전기설비로부터 사고 징후 검출을 위해 전류 모니터를 이용하여 전기설비 전체의 전류를 측정하고 이를 분석하였다. 그 결과, 전류 파형의 특정 영역에서 사고의 진행 정도가 클수록 전류 펄스의 크기가 커지고, 발생 빈도도 증가하였다. 특히 전류파형에 대한 표준편차와 확률분포는 부하의 수와 무관하게 사고단계별로 뚜렷한 차이를 보였다.

• 한국안전학회지
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• 제23권2호
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• pp.65-71
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• 2008

As the interest in traffic safety has been increasing recently, social movement is being made to reduce the number of traffic accidents and the view on improving the mobility of the existing roads is being converted into on establishing traffic safety as a priority. The increase of traffic accidents related to an intersection in a state that traffic accidents are decreasing overall may suggests the necessity to investigate the specific causes. In addition, we have to consider them when establishing the measures against traffic accidents in a intersection by investigating and analyzing the influences and factors that may affect traffic accidents. To induce the accident severity model, we collected the factors that affect accidents and then applied the Poisson Regression Model among nonlinear regression analysis by verifying the distribution of variables. As a result of the analysis, it turned out that the volume of traffic on main roads, the right turn ratio on sub-roads, the number of ways out on sub-roads, the number of exclusive roads for a left turn, the signals for a right turn on main roads, and an intersect angle were the factors that affect the accident severity.

• Journal of Acupuncture Research
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• 제27권2호
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• pp.143-154
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• 2010

Objectives : The main purpose of this research is to make a survey of the effective way of the Oriental medical care and to enlarge the range of Oriental medical treatment about neck pain patients caused by traffic accident. Methods : In following research. we surveyed 197 neck pain patients of traffic accident who admitted in Daegu Hanny University Oriental Medical Hospital from May, 1, 2008 to september, 30, 2009 according to medical chart. Results : 1. In distribution of gender and age, the rate of female was 74.6%, that of male was 25.4% and the majority of the patients were thirties. In results of treatment according to gender and age, the effective rate of female was higher than the male. that of teens was highest. 2. In total results of treatment, the highest distribution was fair with 40.1%, the effective rate was 90.9%. 3. In category of traffic accident patterns, the highest distribution was car to car crash with 81.7%. The results of treatment according to the significance test were shown statistical significance. 4. In admission days, the highest distribution was 6~10 days with 41.1%. The highest effective rate was 21~25days with 100%. The results of treatment according to the significance test were shown statistical significance. Conclusions : This study suggested that Oriental medical treatments are effective in patients of neck pain caused by traffic accident. Further studies are needed for progress report that after accident patients are discharged from the Oriental Medical Hospital.

• Nuclear Engineering and Technology
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• 제47권1호
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• pp.33-46
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• 2015

During the past 10 years, the Korea Atomic Energy Research Institute (KAERI) has performed a study to control hydrogen gas in the containment of the nuclear power plants. Before the Fukushima accident, analytical activities for gas distribution analysis in experiments and plants were primarily conducted using a multidimensional code: the GASFLOW. After the Fukushima accident, the COM3D code, which can simulate a multidimensional hydrogen explosion, was introduced in 2013 to complete the multidimensional hydrogen analysis system. The code validation efforts of the multidimensional codes of the GASFLOW and the COM3D have continued to increase confidence in the use of codes using several international experimental data. The OpenFOAM has been preliminarily evaluated for APR1400 containment, based on experience from coded validation and the analysis of hydrogen distribution and explosion using the multidimensional codes, the GASFLOW and the COM3D. Hydrogen safety in nuclear power has become a much more important issue after the Fukushima event in which hydrogen explosions occurred. The KAERI is preparing a large-scale test that can be used to validate the performance of domestic passive autocatalytic recombiners (PARs) and can provide data for the validation of the severe accident code being developed in Korea.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.663-668
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• 1996

A two-dimensional continuum model for the prediction of the hydrogen mixing phenomena in the containment compartment under the severe accident conditions is developed. The model could predict well the distribution of time-dependent hydrogen concentration for selected HEDL Experiment. For a simulation of these experiments, the hydrogen is mixed uniform over the test compartment. To predict the extent of non-uniform distribution, the dominant factors such as the geometrical shape of obstacle and velocity of source injection in mixing phenomena are investigated. If the obstacle disturbing the flow of gas mixture exists in the compartment, the uniform distribution of hydrogen may be not guaranteed. The convective circulation of gas flow is separately formed up and down of the obstacle position, which makes a difference of hydrogen concentration between the upper and lower region of the compartment. The recirculation flow must have a considerable mass flow rate relative to velocity of the source injection to sustain the well-mixed conditions of hydrogen.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2004년도 추계학술대회 논문집
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• pp.1105-1108
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• 2004

This paper describes the structural analysis result and load test result of accident EMU(Electric Multiple Units). Structural analysis and load test of EMU were performed for the criteria of safety assessment. Structural analysis using commercial I-DEAS software provided important information on the stress distribution and load transfer mechanisms as well as the amount of damages during rolling stock crash. The purpose of the load test is to evaluate a safety which carbody structure shall be considered fully sufficient rigidity so as to satisfy proper system function under maximum load and operating condition. The results have been used to provide the critical information for the criteria of safety assessment.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.886-892
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• 2024

The cladding temperatures and axial mass distribution computed by MAAP5 were compared with their measured values in the test bundle of the Phebus Test FPT0. The computed cladding temperatures were in good agreed with the measured values in the pre-transient phase. In the transient heat-up phase, the computed temperatures were overestimated by the Baker-Just correlation in MAAP5, but the computed temperatures could simulate the subsequently measured values. The computed mass distribution in the axial direction was in qualitative agreement with the measured one for post-test fuel damage observations. The calculated flow areas of inner and outer regions in the test bundle were compared with the photographic observations. MAAP5 computed them at the height of 0.2 m where the molten pool formed was in qualitative agreement with the photographic observations. It was found that the remaining steam flow paths might be caused by the gas-liquid two-phase flow counter-current flow limitation.