• Nuclear Engineering and Technology
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• 제12권2호
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• pp.88-98
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• 1980

원통형 관속을 흐르는 뜨거운 단상 유체의 응고 과정을 해석적인 방법과 실험적인 방법으로 연구 하였다. 파라핀초와 Wood's Metal을 뜨거운 유체로 사용하여 일련의 실험을 하였다. 이 실험 데이터로 부룩해븐 연구소에서 개발한 응고과정에 대한 기존 준정적 수리해석 모델을 증명하였다. 또한 이 실험결과, 수직관속을 순간적으로 흐르며 응고하는 과정에 미치는 여러가지 매개변수의 영향에 관한 자료를 얻게 되었다. 이 실험에 사용한 기구와 실험 방법 도 아울러 기술하였다. 녹은 유체의 순간적으로 흘러내리는 양에 대한 수학적 모델의 예측 결과를 실험데이터와 비교하기 위해 도표로 제시하였다. 또한, 수학적 모델을 고속증식로(LMFBR)에 사고가 일어 났을 경우에 응용하여 보았다.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2008년도 춘계학술대회논문집
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• pp.268-271
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• 2008

If a reactor vessel is failed to retain a molten corium in a postulated severe accident, the molten corium is released outside the reactor vessel into a reactor cavity. The molten corium would attack the concrete wall and basemat of the reactor cavity, which may lead to inevitable concrete decompositions and possible radiological releases. In the OECD/MCCI project, a series of tests were performed to secure the data for cooling the molten corium spread out at the reactor cavity and for the long-term CCI (Core Concrete Interaction). Also, a MCCI (Molten Core Concrete Interaction) analysis code, CORQUENCH was upgraded at Argonne National Laboratory with embedding the new models developed for the tests. This paper deals with analyses of MCCI at plant scale under the conditions of top flooding using the upgraded CORQUENCH code. The modeling approach is briefly summarized first, followed by presentation of a validation calculation that illustrates the predicative capability of the modeling tool. With this background in place, the model is then used to carry out a parametric set of scoping calculations that define approximate coolability envelopes for the LCS (Limestone Common Sand) concrete that has been evaluated in the OECD/MCCI project.

• Nuclear Engineering and Technology
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• 제45권3호
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• pp.295-310
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• 2013

Industry- or regulatory-sponsored research activities on the resolution of Generic Safety Issue (GSI)-191 were reviewed, especially on the chemical effects. Potential chemical effects on the head loss across the debris-loaded sump strainer under a post-accident condition were experimentally evidenced by small-scale bench tests, integrated chemical effects test (ICET), and vertical loop head loss tests. Three main chemical precipitates were identified by WCAP-16530-NP: calcium phosphate, aluminum oxyhydroxide, and sodium aluminum silicate. The former two precipitates were also identified as major chemical precipitates by the ICETs. The assumption that all released calcium would form precipitates is reasonable. CalSil insulation needs to be minimized especially in a plant using trisodium phosphate buffer. The assumption that all released aluminum would form precipitates appears highly conservative because ICETs and other studies suggest substantial solubility of aluminum at high temperature and inhibition of aluminum corrosion by silicate or phosphate. The industry-proposed chemical surrogates are quite effective in increasing the head loss across the debris-loaded bed and more effective than the prototypical aluminum hydroxide precipitates generated by in-situ aluminum corrosion. There appears to be some unresolved potential issues related to GSI-191 chemical effects as identified in NUREG/CR-6988. The United States Nuclear Regulatory Commission, however, concluded that the implications of these issues are either not generically significant or are appropriately addressed, although several issues associated with downstream in-vessel effects remain.

• Nuclear Engineering and Technology
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• 제56권10호
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• pp.4068-4076
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• 2024

In the event of severe accidents in pressurized water reactor (PWR) nuclear power plants, the potential leakage of radioactive aerosols through containment cracks poses a considerable radioactive hazard to the public. Understanding aerosol transport and retention in cracks helps reduce the conservatism and uncertainty of radioactive hazard assessment. Concrete cracks are recognized as a pivotal pathway for the leakage of radioactive aerosols, and several researchers have undertaken experimental investigations concerning the aerosol transport and retention in concrete cracks. However, the majority of these studies have rather low gas flow Reynolds numbers. In this work, an experimental setup is built to study aerosol transport and retention in concrete cracks under high Reynolds number flow. The TiO₂ aerosol with a mass median diameter of 1 ㎛ and two concrete crack specimens are used in experiments. The results of gas flow experiments indicate that the Reynolds number is capable of reaching 10547. Combining the flow experimental data and Suzuki's formula, the equivalent heights of these two crack specimens are approximated as 303.67 ㎛ and 231.48 ㎛. The experimental results indicate a notably high retention rate of aerosols, exceeding 0.8. Furthermore, under high Reynolds number flow, the retention rate varies over a relatively narrow range, with the larger the equivalent height of the crack resulting in a lower retention rate. The experimental results match well with the mechanistic analysis based on inertial deposition theory, demonstrating the rationality of the inertial deposition theory.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회B
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• pp.3491-3496
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• 2007

If a molten core is released from a reactor vessel into a reactor cavity during a severe accident, an important safety issue of coolability of the molten core from top-flooding and concrete ablation due to a molten core concrete interaction (MCCI) is still unresolved. The released molten core debris would attack the concrete wall and basemat of the reactor cavity, which will lead to inevitable concrete decompositions and possible radiological releases. In a OECD/MCCI project scheduled for 4 years from 2002. 1 to 2005. 12, a series of tests were performed to secure the data for cooling the molten core spread out at the reactor cavity and for the 2-D long-term core concrete interaction (CCI). The tests included not only separate effect tests such as a melt eruption, water ingression, and crust failure tests with a prototypic material but also 2-D CCI tests with a prototypic material under dry and flooded cavity conditions. The paper deals with the transient simulations on the CCI-2 test by using a severe accident analysis code, CORQUENCH, which was developed at Argonne National Laboratory (ANL). Similar simulations had been already per for me d by using MELCOR 1.8.5 code. Unlike the MELCOR 1.8.5, the CORQUENCH includes a melt eruption mode I and a newly developed water ingression model based on the water ingression tests under the OECD/MCCI project. In order to adjust the geometrical differences between the CCI-2 test (rectangular geometry) and the simulations (cylindrical geometry), the same scaling methodology as used in the MELCOR simulation was applied. For the direct comparison of the simulation results, the same inputs for the MELCOR simulation were used. The simulation results were compared with the previous results by using MELCOR 1.8.5.

• Journal of Radiation Protection and Research
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• 제41권3호
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• pp.237-244
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• 2016

Background: Korea Atomic Energy Research Institute is developing a fission product transport module for predicting the behavior of radioactive materials in the primary cooling system of a nuclear power plant as a separate module, which will be connected to a severe accident analysis code, Core Meltdown Progression Accident Simulation Software (COMPASS). Materials and Methods: This fission product transport (COMPASS-FP) module consists of a fission product release model, an aerosol generation model, and an aerosol transport model. In the fission product release model there are three submodels based on empirical correlations, and they are used to simulate the fission product gases release from the reactor core. In the aerosol generation model, the mass conservation law and Raoult's law are applied to the mixture of vapors and droplets of the fission products in a specified control volume to find the generation of the aerosol droplet. In the aerosol transport model, empirical correlations available from the open literature are used to simulate the aerosol removal processes owing to the gravitational settling, inertia impaction, diffusiophoresis, and thermophoresis. Results and Discussion: The COMPASS-FP module was validated against Aerosol Behavior Code Validation and Evaluation (ABCOVE-5) test performed by Hanford Engineering Development Laboratory for comparing the prediction and test data. The comparison results assuming a non-spherical aerosol shape for the suspended aerosol mass concentration showed a good agreement with an error range of about ${\pm}6%$. Conclusion: It was found that the COMPASS-FP module produced the reasonable results of the fission product gases release, the aerosol generation, and the gravitational settling in the aerosol removal processes for ABCOVE-5. However, more validation for other aerosol removal models needs to be performed.

• International Journal of Naval Architecture and Ocean Engineering
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• 제11권1호
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• pp.22-32
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• 2019

On offshore platforms, oil and gas leaks are apt to be the initial events of major accidents that may result in significant loss of life and property damage. To prevent accidents induced by leakage, it is vital to perform a case-specific and accurate risk assessment. This paper presents an integrated method of Ddynamic Qquantitative Rrisk Aassessment (DQRA)-using the Decision Making Trial and Evaluation Laboratory (DEMATEL)-Bayesian Network (BN)-for evaluation of the system vulnerabilities and prediction of the occurrence probabilities of accidents induced by leakage. In the method, three-level indicators are established to identify factors, events, and subsystems that may lead to leakage, fire, and explosion. The critical indicators that directly influence the evolution of risk are identified using DEMATEL. Then, a sequential model is developed to describe the escalation of initial events using an Event Tree (ET), which is converted into a BN to calculate the posterior probabilities of indicators. Using the newly introduced accident precursor data, the failure probabilities of safety barriers and basic factors, and the occurrence probabilities of different consequences can be updated using the BN. The proposed method overcomes the limitations of traditional methods that cannot effectively utilize the operational data of platforms. This work shows trends of accident risks over time and provides useful information for risk control of floating marine platforms.

• Journal of Radiation Protection and Research
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• 제44권4호
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• pp.128-148
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• 2019

Massive environmental monitoring has been conducted continuously since the Fukushima Daiichi Nuclear Power accident in March of 2011 by different monitoring methods that have different features together with migration studies of radiocesium in diverse environments. These results have clarified the characteristics of radiological environments and their temporal change around the Fukushima site. At three months after the accident, multiple radionuclides including radiostrontium and plutonium were detected in many locations; and it was confirmed that radiocesium was most important from the viewpoint of long-term exposure. Radiation levels around the Fukushima site have decreased greatly over time. The decreasing trend was found to change variously according to local conditions. The air dose rates in environments related to human living have decreased faster than expected from radioactive decay by a factor of 2-3 on average; those in pure forest have decreased more closely to physical decay. The main causes of air dose rate reduction were judged to be radioactive decay, movement of radiocesium in vertical and horizontal directions, and decontamination. Land-use categories and human activities have significantly affected the reduction tendency. Difference in the air dose rate reduction trends can be explained qualitatively according to the knowledge obtained in radiocesium migration studies; whereas, the quantitative explanation for individual sites is an important future challenge. The ecological half-lives of air dose rates have been evaluated by several researchers, and a short-term half-life within 1 year was commonly observed in the studies. An empirical model for predicting air dose rate distribution was developed based on statistical analysis of an extensive car-borne survey dataset, which enabled the prediction with confidence intervals. Different types of contamination maps were integrated to better quantify the spatial data. The obtained data were used for extended studies such as for identifying the main reactor that caused the contamination of arbitrary regions and developing standard procedures for environmental measurement and sampling. Annual external exposure doses for residents who intended to return to their homes were estimated as within a few millisieverts. Different forms of environmental data and knowledge have been provided for wide spectrum of people. Diverse aspects of lessons learned from the Fukushima accident, including practical ones, must be passed on to future generations.

• 예술인문사회 융합 멀티미디어 논문지
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• 제8권5호
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• pp.417-425
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• 2018

본 논문에서는 조선해양 작업장에서 발생하는 중대재해를 줄이고 체계적인 안전관리를 위하여 ETA(event tree analysis)기반 시나리오 도출 및 ICT 기술접목을 통한 안전관리 구축방안을 제안한다. 안전보건공단과 (구)국민안전처의 통계결과 조선해양 관련 중대재해 중 가장 많이 발생하는 사고유형은 떨어짐이고, 주요사고원인은 안전대 미착용 및 안전대 고리 미체결이다. 이 문제를 해결하기 위해 ETA기반의 시나리오를 작성하여 안전사항에 따른 결과를 도출하고 이 결과를 바탕으로 사고예방을 위한 ICT 기술접목으로 해결방안을 제시한다. ETA 기반 시나리오 도출 및 ICT 기술접목을 통해 제안한 해결방안으로는 안전대 및 안전모 착용여부 감지시스템, 안전대 고리 체결여부 감지시스템, 안전거리 측정을 위한 걸이설비 측정시스템이다. 안전사항별 시스템을 통해 작업자의 떨어짐 위험을 줄여 사망확률을 낮출 수 있다. 제안한 방안을 통해 사고를 예방함으로써 조선해양 분야의 중대재해를 줄이고 체계적인 안전관리를 도모한다.

• 한국산업보건학회지
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• 제30권2호
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• pp.99-108
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• 2020

Objectives: This study aimed to compare the regulatory systems for laboratory safety and health management between Korea and Germany and discuss the implications. Methods: Laboratory safety and health regulations for legal enforcement and relevant technical guidelines in Korea and Germany were reviewed. Results: Lab safety and health management is enforced by the Act on the Establishment of Safe Laboratory Environment in Korea. Most provisions focus on supervisory control, that is, the principal's liability is emphasized. In addition, there is a lack of laboratory-specific procedures for safety and health management in the act since it is stipulated that other relevant regulations apply to some technical contents. Non-compulsory technical guidelines for lab safety and health management are also provided by the Korea Occupational Safety and Health Agency (KOSHA) in order to enable researchers to follow safe procedures. There is no independent regulation for lab safety and health in Germany, and it is also governed by several regulations. The German Social Accident Insurance Institute provides technical guidelines on lab safety and health, and these contain more specific content to allow them to be followed more easily compared to the KOSHA guidelines. The most remarkable differences between the regulation of each country were contents of the risk assessment and specific protect measures from hazardous agents. Conclusions: Regulatory control is an essential way to prevent accidents, but it is more important to create an environment in which all stakeholders, including individual lab members, are allowed to participate actively in safety and health management activities.