• 대한안전경영과학회지
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• 제20권3호
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• pp.27-36
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• 2018

The steel pipe manufacturing industry deals with facilities and materials. Especially thermal facilities are close to vapor cloud explosion (VCE) and may cause secondary damage to facilities because they deal with corrosive substances such as hydrofluoric acid, sulfuric acid and acid, fire, explosion, leakage etc. It is in danger. In this study, hazard identification method was conducted using HAZOP techniques and quantitative risk analysis was conducted using e-CA, a program that supports accident impact analysis. Equipment in the influence range of ERPG - 3 was determined to be a facility requiring replacement. It was decided that neutralization is necessary using slaked lime. Based on the cost of loss, We presented the proper replacement which is the timing of the dangerous facility. As a result, It was ideal to replace the facilities with 20 years of heat treatment facilities, one year of hydrofluoric acid storage tank, 20 years of sulfuric acid storage tank, and 5 years of hydrochloric acid storage tank.

• Nuclear Engineering and Technology
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• 제21권1호
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• pp.1-11
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• 1989

원자로 정지동안에도, 잔열제거계통은 그 기능이 계속 유지되어야 하나, 실제로 가압 경수로에서 냉각상실고가 많이 발생되어 있다. 본 논문은 원자로 정지중의 냉각기능상실을 예방하고, 또한 냉각기능상실로 인한 노심손상의 중대성을 완화시키기 위한 대책을 강구하기 위한 시도로서, 전형적인 가압경수로에 대한 사고/고장 수목과 운전원실수 확률을 위한 HCR 모델, 초기 사상의 빈도를 위한 2단계 bayesian 방법 및 고장난 계통의 회복 활률을 위한 계단함수 모델 등을 이용한 원자로 정지 위해도 모델을 개발하여, 잔열제거계통의 신뢰도를 분석하였다. 그 결과는 원자로가 정지 중일 때의 위해도가 운전중일 때 이것에 비해 별로 낮지 않은 것으로 나타났으며, 몇 가지의 설계개선을 통하여 냉각기능상실로 인한 노심 손상확률을 상당히 낮출 수 있는 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제48권2호
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• pp.457-469
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• 2016

During the late phase of severe accidents of light water reactors, a porous debris bed is expected to develop on the bottom of the flooded reactor cavity after breakup of the melt in water. The geometrical configuration, i.e., internal and external characteristics, of the debris bed is significant for the adequate assessment of the coolability of the relocated corium. The internal structure of a debris bed was investigated experimentally using the DAVINCI (Debris bed research Apparatus for Validation of the bubble-Induced Natural Convection effect Issue) test facility. Particle sedimentation under the influence of a two-phase natural convection flow due to the decay heat in the debris bed was simulated by dropping various sizes of particles into a water vessel with air bubble injection from the bottom. Settled particles were collected and sieved to obtain the particle mass, size distribution in the radial and axial positions, and the bed porosity and permeability. The experimental results showed that the center part of the particle bed tended to have larger particles than the peripheral area. For the axial distribution, the lower layer had a higher fraction of larger particles. As the sedimentation progressed, the size distribution in the upper layers can shift to larger sizes because of the higher vapor generation rate and stronger flow intensity.

• Nuclear Engineering and Technology
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• 제22권1호
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• pp.12-19
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• 1990

1981년 6월 9일 고리 1호기 원자력발전소에서 발생한 외부 전원 상실사고 자료를 근거로 RELAP5/MOD2코드모델 평가를 하였다. 계산된 주요 열ㆍ수력학 변수를 실측자료와 비교 분석하였으며 증기발생기의 Nodalization 민감도 분석이 수행되었다. 계산된 열ㆍ수력학 변수는 실측치와 비교적 잘 일치하고 있으며, 이러한 유형의 사고 분석에 RELAP5/MOD2가 적합하다는 것을 보였다. 그러나 가압기 압력과 수위변동에서는 상당한 차이를 보였으며 높게 계산되었다. 이러한 사실은 RELAP5의 수직관에서의 층류 열전달 모델에 기인하는 것으로 해당모델의 개선을 요하고 있다는 것을 알았다. 그리고 증기발생기의 Nodalization 연구를 통하여 수위변동을 잘 예측하기 1위해서는 증기발생기 증기 Dome와 Downcomer사이에 압력을 전달시켜주는 유로를 모델링 하여야 한다는 것을 알았다.

• Korean Chemical Engineering Research
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• 제57권1호
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• pp.28-41
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• 2019

본 연구는 소듐냉각고속로 원형로 소듐-물 반응 압력완화계통의 성능 해석을 목적으로 한다. 증기발생기의 전열관 파단에 의한 대규모 물 누출 사고 발생 시, 증기발생기 전열관 내측의 물을 급수덤프탱크로 배출하고 전열관 외측의 소듐 및 반응생성물을 소듐덤프탱크로 배출 할 때 유체의 거동을 해석하여 계통 설계요건의 적절성을 평가하였다. 증기발생기 쉘 측의 액체와 중간열전달계통 내 소듐이 모두 배출되는데 소요되는 시간은 약 50초이고, 증기발생기 전열관 측의 급수가 모두 배출되는데 소요되는 시간은 약 2.5초로 계산되었다. 증기발생기와 중간열전달계통 내 유체가 덤프탱크로 배출되는 동안 전열관 측의 압력은 쉘 측의 압력보다 높게 유지되어 쉘 측의 소듐이 전열관 측으로 역류하는 현상은 없는 것으로 해석되었다. 본 연구의 결과는 SFR 원형로 소듐-물반응압력완화계통의 성능 평가에 대한 기초자료로 활용할 예정이다.

• 시스템엔지니어링학술지
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• 제16권2호
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• pp.97-109
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• 2020

On March 11, 2011, an earthquake followed by a tsunami caused an extended station blackout (SBO) at the Fukushima Dai-ichi NPP Units. The accident was initiated by a total loss of both onsite and offsite electrical power resulting in the loss of the ultimate heat sink for several days, and a consequent core melt in some units where proper mitigation strategies could not be implemented in a timely fashion. To enhance the plant's coping capability, the Diverse and Flexible Strategies (FLEX) were proposed to append the Emergency Operation Procedures (EOPs) by relying on portable equipment as an additional line of defense. To assess the success window of FLEX strategies, all sources of uncertainties need to be considered, using a physics-based model or system code. This necessitates conducting a large number of simulations to reflect all potential variations in initial, boundary, and design conditions as well as thermophysical properties, empirical models, and scenario uncertainties. Alternatively, data-driven models may provide a fast tool to predict the success window of FLEX strategies given the underlying uncertainties. This paper explores the applicability of Artificial Intelligence (AI) to identify the success window of FLEX strategy for extended SBO. The developed model can be trained and validated using data produced by the lumped parameter thermal-hydraulic code, MARS-KS, as best estimate system code loosely coupled with Dakota for uncertainty quantification. A Systems Engineering (SE) approach is used to plan and manage the process of using AI to predict the success window of FLEX strategies under extended SBO conditions.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2311-2320
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• 2022

The main outcomes of the experiments H2P6 performed in the thermal-hydraulics large-scale PANDA facility at PSI in the frame of the OECD/NEA HYMERES-2 project are presented in this article. The experiments of the H2P6 series consists of two PANDA tests characterized by the activation of three (H2P6_1) or one (H2P6_2) cooler(s) in an initially stratified and pressurized containment atmosphere. The initial stratification is defined by a helium-rich region located in the upper part of the vessel and a steam/air atmosphere in the lower part. The activation of the cooler(s) results i) in the condensation of the steam in the vicinity of the cooler(s), ii) the corresponding activation of large scale natural circulation currents in the vessel atmosphere, with the result of iii) the re-distribution and mixing of the Helium stratification initially located in the upper half of the vessel and iv) the continuous pressure decay. The initial helium layer represents hydrogen generated in a postulated severe accident. The main question to be answered by the experiments is whether or not the interaction of the different, localized cooler units would be important for the application of numerical methods. The paper describes the initial and boundary conditions and the experimental results of the H2P6 series with the suggestion of simple scaling laws for both experiments in terms of i) the temperature difference(s) across the cooler(s), ii) the transient steam and helium content and iii) the pressure decay in the vessel. The outcomes of this scaling indicate that the interaction between separate, closely localized units does not play a prominent role for the present experiments. It is therefore reasonable to model several units as one large component with equivalent heat transfer area and total water flow rate.

• 한국기계가공학회지
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• 제20권6호
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• pp.33-43
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• 2021

A60 class bulkhead penetration piece is a fire resistance system installed on a bulkhead compartment to protect lives and to prevent flame diffusion in a fire accident on a ship and offshore plant. This study focuses on the approximate optimization of the fire resistance design of the A60 class bulkhead penetration piece using a multi-island genetic algorithm. Transient heat transfer analysis was performed to evaluate the fire resistance design of the A60 class bulkhead penetration piece. For approximate optimization, the bulkhead penetration piece length, diameter, material type, and insulation density were considered discrete design variables; moreover, temperature, cost, and productivity were considered constraint functions. The approximate optimum design problem based on the meta-model was formulated by determining the discrete design variables by minimizing the weight of the A60 class bulkhead penetration piece subject to the constraint functions. The meta-models used for the approximate optimization were the Kriging model, response surface method, and radial basis function-based neural network. The results from the approximate optimization were compared to the actual results of the analysis to determine approximate accuracy. We conclude that the radial basis function-based neural network among the meta-models used in the approximate optimization generates the most accurate optimum design results for the fire resistance design of the A60 class bulkhead penetration piece.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.71-79
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• 2023

Air ingress incidents are major safety accidents in very-high-temperature reactors (VHTRs). Air containing a high volume fraction of oxygen may cause severe oxidation of core components at the VHTR, especially for the significantly thin alloy tube wall in the intermediate heat exchanger (IHE). The research objects of this study are Inconel 617 and Incoloy 800H, two candidate alloys for IHE in VHTR. The air ingress accident scenario is simulated with high-temperature air flow at 950 ℃. A continuous oxide scale was formed on the surfaces of both the alloys after the experiment. Because the oxide scale of Inconel 617 has a loose structure, whereas that of Incoloy 800H is denser, Inconel 617 exhibited significantly more severe internal oxidation than Incoloy 800H. Further, Inconel 617 showed a significant decrease in ultimate tensile strength and plasticity after aging for 200 h, whereas Incoloy 800H maintained its tensile properties satisfactorily. Through control experiment under vacuum, we preliminarily concluded that serious internal oxidation is the primary reason for the decline in the tensile properties of Inconel 617.

• 한국압력기기공학회 논문집
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• 제18권2호
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• pp.37-42
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• 2022

An influence analysis on multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout is performed to compare the plant responses according to the number of ruptured u-tubes under the assumption of a total of 10 ruptured u-tubes. In all calculation cases, the transient behaviour of major thermal-hydraulic parameters, such as the discharge flow rate through the ruptured u-tubes, reactor header pressure, and void fraction in the fuel channels is found to be overall similar to that of the base case having a single SG with 10 u-tubes ruptured. Additionally, as the conditions of low-flow coolant with high void fraction in the broken loop continued, causing the degradation of decay heat removal, the peak cladding temperature (PCT) would be expected to exceed the limit criteria for ensuring nuclear fuel integrity. However, despite the same total number of ruptured u-tubes, because of the different connection configuration between the SG and pressurizer, a difference is foud in time between the pressurizer low-level signal and reactor header low-pressure signal, affecting the time to trip the reactor and to reach the PCT limit. The present study is expected to provide the technical basis for the accident management strategy for mSGTR transient conditions of CANDU-6 plants.