• Nuclear Engineering and Technology
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• 제56권2호
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• pp.636-643
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• 2024

The coolability of the debris bed with a simulant of solidified corium is experimentally studied, focusing on the effects of the structure of the axial stratified debris bed on the dryout heat flux (DHF). DHF was obtained for the four structures with different particle sizes for the axial stratified debris bed under top flooding. The experimental results show that the dryout position of the axial stratified debris bed is formed at the stratified interface indicated by the temperature rise, and the DHF of the axial stratified bed is much lower than that of the homogeneous bed packed with the upper small particles. To predict the dryout heat flux of the stratified debris beds, by considering the properties of the mixed area, a one-dimensional dryout heat flux model of the porous medium is derived from a water and vapor momentum equation for porous medium, two-phase permeability modifications, interfacial drag, and the correlation between capillary pressure and liquid saturation and verified with the experimental data. The modified model can give reasonable results under different structures.

• 한국가스학회지
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• 제25권6호
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• pp.35-44
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• 2021

반도체 공정 중 PR (Photo Resist) 공정은 여러 인화성 물질을 혼합하여 사용하는 공정으로, 공정 장비가 Clean Room 내 설치되어 인화성 물질이 누출되는 경우 질식, 화재, 폭발 사고로 이어질 위험성이 크므로, 물질 누출시 발생할 수 있는 사고에 대한 영향을 분석하고, 작업자들의 안전을 보장할 수 있는지 평가하는 것이 필요하다. 본 연구는 FLACS CFD - Fire Module을 이용하여 10개의 누출, 화재 시나리오에 대한 CFD Simulation을 통해 Clean Room 내부 가상으로 설정한 Monitor Point에서의 복사열 및 온도 변화 값을 확인하였다. Clean Room 내부에서 발생한 화재는 층간 구조물에 높은 복사열을 전달하지만, 그 범위가 상당히 제한적이며, 단일 화재 사고로 인해 붕괴될 가능성이 희박하다. 화재 사고로 인해 탈출구로 이어지는 2 곳의 계단이 동시에 높은 복사열에 노출되는 시나리오는 없어, 화재 발생시 작업자가 탈출 가능하였다. 또한 복사열 및 온도 상승의 수준은 아래층으로 이동하면서 급격하게 낮아지는 것을 확인하였으며, API 520의 기준에 따라 작업자가 30초 동안 버틸 수 있는 복사열인 6.31 kW/m²에 노출된 작업자도 화재가 발생한 Clean Room 내부에서 충분히 탈출할 수 있음을 확인하였다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.713-718
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• 1997

In the TMI-2 accident, approximately twenty(20) tons of molten core material drained into the lower plenum. Early advanced light water reactor (LWR) designs assumed a lower head failure and incorporated various measures for ex-vessel accident mitigation. However, one of the major findings from the TMI-2 Vessel Investigation Project was that one part of the reactor lower head wall estimated to have attained a temperature of 1100$^{\circ}C$ for about 30 minutes has seemingly experienced a comparatively rapid cooldown with no major threat to the vessel integrity. In this regard, recent empirical and analytical studies have shifted interests to such in-vessel retention designs or strategies as reactor cavity flooding, in-vessel flooding and engineered gap cooling of the vessel Accurate thermohydrodynamic and creep deformation modeling and rupture prediction are the key to the success in developing practically useful in-vessel accident/risk management strategies. As an advanced in-vessel design concept, this work presents the COrium Attack Syndrome Immunization Structures (COASIS) that are being developed as prospective in-vessel retention devices for a next-generation LWR in concert with existing ex-vessel management measures. Both the engineered gap structures in-vessel (COASISI) and ex-vessel (COASISO) are demonstrated to maintain effective heat transfer geometry during molten core debris attack when applied to the Korean Standard Nuclear Power Plant(KSNPP) reactor. The likelihood of lower head creep rupture during a severe accident is found to be significantly suppressed by the COASIS options.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2504-2515
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• 2023

In past years the Paul Scherrer Institute (PSI, Switzerland) and the Karlsruhe Institue of Technology (KIT, Germany)) collaborated to develop a model to account for the active role of nitrogen in the air oxidation of a Zircalloy cladding. The "PSI-KIT Nitriding Model for Zirconium based Fuel Cladding" model was implemented at PSI into PSI-MELCOR 1.8.6. In order to make a preliminary evaluation of the effect of the new model on the evolution of full-scale spent fuel pool accidents, one spent fuel pool event was analyzed using the PSI research version of PSI-MELCOR 1.8.6, which includes the nitriding model. To adapt an existing input deck for the calculations, a sensitivity study was conducted to find an optimal nodalization for the analyses. The nitriding model results were compared to those calculated with the MELCOR 1.8.6-PSI without the new nitriding model. The results demonstrate the effect of the nitriding reactions in spent fuel pool accident progression. Moreover, they confirm the impact of ZrN formation during cladding oxidation in air when the oxidation reactions lead to oxygen starvation inside the fuel assemblies. The nitriding reaction led to higher chemical heat generation during the accident and to an earlier failure of the cladding than when the effect of nitrogen reactions was not considered. It should be noted that the nitriding model, as implemented in the PSI version of MELCOR 1.8.6 has not yet been conclusively validated. Thereby the results presented in this paper should be treated as a preliminary demonstration of the capabilities of the model.

• 대한기계학회논문집B
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• 제25권9호
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• pp.1237-1244
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• 2001

An experimental study was carried out to identify the various regimes of natural convective pool boiling and to determine the boiling heat transfer curve and Critical Heat Flux(CHF) on a vertical square surface having a 70mm width and a 70mm height. The heater made of copper block with embedded cartridge heaters is submerged in a water tank at atmospheric pressure. As the heat flux increases from 100kW/㎡ to 1.2MW/㎡, the heat transfer regime migrates from the nucleate boiling to the film boiling. The boiling heat transfer data are fitted by Rohsenow type correlation. An explosive vapor generation on the heated surface, whose size and frequency are characterized by the heat flux, is visualized using a high speed digital imaging system.

• 한국재료학회지
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• 제21권7호
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• pp.364-370
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• 2011

Currently, there are two main issues regarding the development of core technologies in the automotive industry: the development of environmentally friendly vehicles and securing a high level of safety in the event of an accident. As part of the efforts to address these issues, research into alternative materials and new car body manufacturing and assembly technologies is necessary, and this has been carried out mainly by the automotive industries. Large press molds for producing car body parts are made of cast iron. With the increase of automobile production and various changes of design, the press forming process of car body parts has become more difficult. In the case of large press molds, high hardness and abrasive resistance are needed. To overcome these problems, we attempted to develop a combined heat treatment process consisting of local laser heat treatment followed by plasma nitriding, and evaluated the characteristics of the proposed heat treatment method. From the results of the experiments, it has been shown that the maximum surface hardness is 864 Hv by the laser heat treatment, 953 Hv by the plasma nitriding, and 1,094 Hv by the combined heat treatment. It is anticipated that the suggested combined heat treatment can be used to evaluate the durability of press mold.

• Nuclear Engineering and Technology
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• 제14권4호
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• pp.186-195
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• 1982

SYSRAN code를 사용하여 고리 1호기의 중기배관파열사고를 분석하였다. SYSRAN code는 중성자출력과 열선속계산은 각각 점근사 중성자 운동방정식과 집중정수 모형을 이용하고 냉각수 계통 과도현상에 대해서는 전 계통을 균일한 압력으로 취급하여 질량 및 에너지 평형방정식을 이용하여 계산한다. 사고 결과를 심각하게 만드는 노심상태로 부냉각재 온도계수가 커지는 노심말기와 증기발생기의 유체함량이 가장 많은 고온 정지상태를 호기조건으로 하여, 격납용기외부의 가장 큰 배관면적인 1.4f $t^2$ 크기의 증기배관이 파열되었을때 Moody critical flow model에 따라 증기가 방출된다고 가정하여 분석하였다. 그 결과 노심의 최대 열선속은 사고후 60초에 정상상대의 38%로서 FSAR의 26%에 비해 높은 값을 나타냈으나 모든 과도현상의 경향은 FSAR의 결과와 잘 일치하였다. 민감도 조사결과 이 사고는 냉각재밀도 계수와 노심 하부공간혼합인자에 가장 민감한 것으로 나타났다. B bank중 한 개의 RCCA가 완전인출 상태에서 노심에 삽입되지 않았다고 가정했을 경우의 FSAR 분석결과인 $F_{$\Delta$H}$를 3.66으로 Fz를 1.55로 하여 DNBR을 계산해 본 결과, 최소 DNBR은 1.62가 되어 핵연료의 손상은 예상되지 않았다. 점근사중성자 운동방정식, 집중 정수모형 및 질량과 에너지평형 방정식을 이용한 계통 과도 현상모델은 발전소 전 계통의 과도 현상의 경향을 연구하는데 적합한 것으로 밝혀졌다.구하는데 적합한 것으로 밝혀졌다.

• 한국철도학회논문집
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• 제12권1호
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• pp.25-30
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• 2009

본 연구에서는 철도 화재 시 구난역에서의 화재 연기의 거동을 파악하기 위하여 상용코드를 사용하여 수치해석 하였다. 화원의 모사와 화재로 인한 생성물의 거동을 예측하기 위해 stoichiometric상태에서 연료 소모량에 따른 연소생성물의 생성률과 산소 소모율을 VHS 모델에 적용하고 종의 보존 방정식을 해석하는 HVHS 모델을 이용하였다. 해석결과 화재 연기는 온도에 따른 밀도 차에 의해 터널의 천장을 따라 이동 하였으며 열원으로부터 멀어지면서 하강하는 형태를 확인 할 수 있었다. 또한 터널 내 공기는 화원으로 집중되었으며 비사고 터널과 사고 터널의 압력 차에 의해 화재연기는 별다른 환기 시스템 없이도 비사고 터널로 유입되지 않았다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 춘계학술발표회요약집
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• pp.205-205
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• 2001

• 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.