• 대한전기학회:학술대회논문집
• /
• 대한전기학회 1998년도 추계학술대회 논문집 학회본부A
• /
• pp.386-392
• /
• 1998

Previous counter-current flow limitation (CCFL) and critical heat flux (CHF) studies included investigations on the inlet entrance, inclined channel and gap effects for the most part. In this study, the local CHF correlation was presented to be used in the numerical analysis for the 3 dimensional hemispherical geometry. Also, first-principle analyses were performed to determine the Maximum heat removal capability from the debris through the gap that may be formed during a core melt accident. The maximum heat removal capability by gap cooling can be applied in quantitatively assessing the severe accident management measures.

• Nuclear Engineering and Technology
• /
• 제45권4호
• /
• pp.459-468
• /
• 2013

During a station blackout (SBO), the initiating event is a loss of Class IV and Class III power, causing the loss of the pumps, used in systems such as the primary heat transporting system (PHTS), moderator cooling, shield cooling, steam generator feed water, and re-circulating cooling water. The reference case of the SBO case does not credit any of these active heat sinks, but only relies on the passive heat sinks, particularly the initial water inventories of the PHTS, moderator, steam generator secondary side, end shields, and reactor vault. The reference analysis is followed by a series of sensitivity cases assuming certain system availabilities, in order to assess their mitigating effects. This paper also establishes the strategies to mitigate SBO accidents. Current studies and strategies use the computer code of the Integrated Severe Accident Analysis Code (ISAAC) for Wolsong plants. The analysis results demonstrate that appropriate strategies to mitigate SBO accidents are established and, in addition, the symptoms of the SBO processes are understood.

• Nuclear Engineering and Technology
• /
• 제50권3호
• /
• pp.356-367
• /
• 2018

A series of tests dedicated to station blackout (SBO) accident scenarios have been recently performed at the $Prim{\ddot{a}}rkreislauf-Versuchsanlage$ (primary coolant loop test facility; PKL) facility in the framework of the OECD/NEA PKL-3 project. These investigations address current safety issues related to beyond design basis accident transients with significant core heat up. This work presents a detailed analysis using the best estimate thermal-hydraulic code TRACE (v5.0 Patch4) of different SBO scenarios conducted at the PKL facility; failures of high- and low-pressure safety injection systems together with steam generator (SG) feedwater supply are considered, thus calling for adequate accident management actions and timely implementation of alternative emergency cooling procedures to prevent core meltdown. The presented analysis evaluates the capability of the applied TRACE model of the PKL facility to correctly capture the sequences of events in the different SBO scenarios, namely the SBO tests H2.1, H2.2 run 1 and H2.2 run 2, including symmetric or asymmetric secondary side depressurization, primary side depressurization, accumulator (ACC) injection in the cold legs and secondary side feeding with mobile pump and/or primary side emergency core coolant injection from the fuel pool cooling pump. This study is focused specifically on the prediction of the core exit temperature, which drives the execution of the most relevant accident management actions. This work presents, in particular, the key improvements made to the TRACE model that helped to improve the code predictions, including the modeling of dynamical heat losses, the nodalization of SGs' heat exchanger tubes and the ACCs. Another relevant aspect of this work is to evaluate how well the model simulations of the three different scenarios qualitatively and quantitatively capture the trends and results exhibited by the actual experiments. For instance, how the number of SGs considered for secondary side depressurization affects the heat transfer from primary side; how the discharge capacity of the pressurizer relief valve affects the dynamics of the transient; how ACC initial pressure and nitrogen release affect the grace time between ACC injection and subsequent core heat up; and how well the alternative feeding modes of the secondary and/or primary side with mobile injection pumps affect core quenching and ensure stable long-term core cooling under controlled boiling conditions.

• Nuclear Engineering and Technology
• /
• 제54권12호
• /
• pp.4514-4521
• /
• 2022

In this study, the effects of heat and radiation on the degradation behaviour of fluoroelastomer under simulated normal operation and a severe accident environment were investigated using sequential testing of gamma irradiation and thermal degradation. Tensile properties and Shore A hardness were measured, and thermogravimetric analysis was used to evaluate the degradation behaviour of fluoroelastomer. Fourier transform infrared spectroscopy and X-ray photoelectron spectroscopy were used to characterize the structural changes of the fluoroelastomer. Heat and radiation generated in nuclear power plant break and deform the chemical bonds, and fluoroelastomer exposed to these environments have decreased C-H and functional groups that contain oxygen and double bonds such as C-O, C=O and C=C were generated. These functional groups were formed by auto oxidation by reacting free radicals generated from the cleaved bond with oxygen in the atmosphere. In this auto oxidation reaction, crosslinks were generated where bonded to each other, and the mobility of molecules was decreased, and as a result, the fluoroelastomer was hardened. This hardening behaviour occurred more significantly in the severe accident environment than in the normal operation condition, and it was found that thermal stability decreased with the generation of unstable structures by crosslinking.

• Nuclear Engineering and Technology
• /
• 제45권6호
• /
• pp.759-766
• /
• 2013

As passive safety features for nuclear power plants receive increasing attention, various studies have been conducted to develop safety systems for 3rd-generation (GEN-III) nuclear power plants that are driven by passive systems. The Passive Auxiliary Feedwater System (PAFS) is one of several passive safety systems being designed for the Advanced Power Reactor Plus (APR+), and extensive studies are being conducted to complete its design and to verify its feasibility. Because the PAFS removes decay heat from the reactor core under transient and accident conditions, it is necessary to evaluate the heat removal capability of the PAFS under hypothetical accident conditions. The heat removal capability of the PAFS is strongly dependent on the heat transfer at the condensate tube in Passive Condensation Heat Exchanger (PCHX). To evaluate the model of heat transfer coefficient for condensation, the Multi-dimensional Analysis of Reactor Safety (MARS) code is used to simulate the experimental results from PAFS Condensing Heat Removal Assessment Loop (PASCAL). The Shah model, a default model for condensation heat transfer coefficient in the MARS code, under-predicts the experimental data from the PASCAL. To improve the calculation result, The Thome model and the new version of the Shah model are implemented and compared with the experimental data.

• Nuclear Engineering and Technology
• /
• 제44권6호
• /
• pp.597-610
• /
• 2012

The passive auxiliary feedwater system (PAFS) is one of the advanced safety features adopted in the APR+, which is intended to completely replace the conventional active auxiliary feedwater system. With an aim of validating the cooling and operational performance of PAFS, an experimental program is in progress at KAERI, which is composed of two kinds of tests; the separate effect test and the integral effect test. The separate effect test, PASCAL ($\underline{P}$AF$\underline{S}$ $\underline{C}$ondensing Heat Removal $\underline{A}$ssessment $\underline{L}$oop), is being performed to experimentally investigate the condensation heat transfer and natural convection phenomena in PAFS. A single, nearly-horizontal U-tube, whose dimensions are the same as the prototypic U-tube of the APR+ PAFS, is simulated in the PASCAL test. The PASCAL experimental result showed that the present design of PAFS satisfied the heat removal requirement for cooling down the reactor core during the anticipated accident transients. The integral effect test is in progress to confirm the operational performance of PAFS, coupled with the reactor coolant systems using the ATLAS facility. As the first integral effect test, an FLB (feedwater line break) accident was simulated for the APR+. From the integral effect test result, it could be concluded that the APR+ has the capability of coping with the hypothetical FLB accident by adopting PAFS and proper set-points of its operation.

• 한국가스학회지
• /
• 제3권3호
• /
• pp.45-50
• /
• 1999

도심지내에 위치한 부천 LPG충전소 사고 조사를 통하여 가장 피해효과가 큰 탱크로리 폭발에 따른 결과를 분석하였다. 분석범위는 BLEVE 현상에 의한 방출열과 과압이 충전소주변에 위치한 구조물이나 인체에 미치는 영향을 대상으로, 실제 현장조사를 통하여 수집된 피해결과와 이론적인 모델(PHAST-Process Hazad Analysis Sortware Tools) 분석 결과를 비교하였다. 부천 LPG 충전소 폭발 사고의 피해효과는 방출열의 경우 두 가지 모두 큰 차이를 보이지 않았으나, 과압의 경우, 실제 피해는 이론적 모델분석결과의 약 $15\%$정도에 해당하는 축소된 거리에서 나타났다. 또한, 충전소 인근에 위치한 구조물에 대한 피해효과는 부분적으로 과압에 의한 균열 및 붕괴 현상보다는 열 효과에 의한 콘크리트 강도 저하와 성상변화가 크게 나타났다.

• 한국안전학회지
• /
• 제26권6호
• /
• pp.13-19
• /
• 2011

The purpose of this study is to examine the cause of damage to electrolytic capacitors and to present the heat generation mechanism in order to prevent the occurrence of similar problems. From the analysis results of electrolytic capacitors collected from accident sites, the fire causing area can be limited to the primary power supply for the initial accident. From the tests performed by applying overvoltage, surge, etc., it is thought that the fuse, varistor, etc., are not directly related to the accidents that occurred. The analysis of the characteristics using a switching regulator showed that the charge and discharge characteristics fell short of standard values. In addition, it is thought that heated electrolytic capacitors caused thermal stress to nearby resistances, elements, etc. It can be seen that the heat generation is governed by the over-ripple current, application of AC overvoltage, surge input, internal temperature increase, defective airtightness, etc. Therefore, when designing an electrolytic capacitor, it is necessary to comprehensively consider the correct polarity arrangement, appropriate voltage application, correct connection of equivalent series resistance(ESR) and equivalent series inductance(SEL), rapid charge and discharge control, sufficient margin of dielectric tangent, etc.

• 한국수소및신에너지학회논문집
• /
• 제30권6호
• /
• pp.578-584
• /
• 2019

The occurrence of white plume in the cooling tower is phenomenon that the steam in the air through the cooling tower fan is condensed again by the cold ambient air to become saturated moist air. Accordingly, this can cause many problems like spoiling landscape around the cooling tower, odor of ambient air, falling accident by frozenness in the winter, and traffic accident, etc. This study was to install the heat exchanger in the inside of the cooling tower in order to prevent the white plume phenomenon in the cooling tower without affecting the performance of cooling tower. In addition, this study was to discharge the part of cooling water into the atmosphere through the recirculation of heat exchanger after creating dry air by heating the saturated moist air to the dew point temperature. At that time, this study was to conduct the experimental study in order to secure the optimal design data to prevent the white plume in the cooling tower because it checked the dry·moist temperature and relative humidity in the inside·outside of cooling tower on the moist air, and evaluated the performance of the heat exchanger.

• Nuclear Engineering and Technology
• /
• 제55권3호
• /
• pp.1125-1139
• /
• 2023

The system-integrated modular advanced reactor 100 (SMART100), an integral-type pressurized water small modular reactor, is based on a novel design concept for containment cooling and radioactive material reduction; it is known as the containment pressure and radioactivity suppression system (CPRSS). There is a passive cooling system using a condensation with non-condensable gas in the SMART CPRSS. When a design basis accident such as a small break loss of coolant accident (SBLOCA) occurs, the pressurized low containment area (LCA) of the SMART CPRSS leads to steam condensation in an incontainment refuelling water storage tank (IRWST). Additionally, the steam and non-condensable gas mixture passes through the CPRSS heat exchanger (CHX) submerged in the emergency cooldown tank (ECT) that can partially remove the residual heat. When the steam and non-condensable gas mixture passes through the CHX, the non-condensable gas can interrupt the condensation heat transfer in the CHX and it degrades CHX performance. In this study, condensation heat transfer experiments of steam and non-condensable gas mixture in the natural circulation loop were conducted. The pressure, temperature, and effects of the non-condensable gas were investigated according to the constant inlet steam flow rate with non-condensable gas injections in the loop.