• 대한기계학회논문집B
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• 제35권4호
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• pp.425-429
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• 2011

원자로의 노심 손상에 따른 노심 용융물의 노외 유출시 코어캐처라고 불리는 설비를 통해 용융물을 억제하고 냉각시키게 된다. 이 때 노외 노심용융물의 거동은 희생물질과의 반응을 포함한 복잡한 물리적, 화학적 현상에 의해 결정된다. 이 연구는 기존의 용융물 거동 실험결과에 대해 용융물의 유동과 열전달의 세부적인 특성을 상용코드를 이용해 해석하여 검증함으로써 코어캐처의 설계에 활용할 수 있도록 하기 위한 것이다. 단순화된 채널에서 시간에 따른 용융물과 공기의 이상유동과 복사열전달을 VOF 모델과 구분종좌법을 적용하여 비정상상태에서 해석한 결과, 열전달에 따른 용융물 내부의 온도 변화 및 이에 따른 점성 변화 등을 예측할 수 있음을 확인하였다. 이러한 접근방식을 기초로 향후 용융물의 조성, 유량 및 용도 등의 조건에 따른 용융물의 거동에 대한 자세한 평가가 필요하다.

• Nuclear Engineering and Technology
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• 제55권6호
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• pp.2069-2087
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• 2023

In nuclear power plant (NPP) accidents, the containment is subject to high temperatures and high internal pressures, which may further trigger serious chain accidents such as core meltdown and hydrogen explosion, resulting in a significantly higher accident level. Therefore, studying the mechanical performance of a containment under high temperature and high internal pressure is relevant to the safety of NPPs. Based on similarity principles, the 1:3.2 scale model of a prestressed concrete containment vessel (PCCV) of a NPP was designed. The loading method, which considers the thermal-pressure coupling conditions, was used. The mechanical response of the PCCV was investigated with a simultaneous increase in internal pressure and temperature, and the failure mechanism of the PCCV under thermal-pressure coupling conditions was revealed.

• 한국유체기계학회 논문집
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• 제17권6호
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• pp.109-114
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• 2014

After the Fukushima accident in 2011, it was revealed that nuclear power plant has the vulnerability to SBO accident and its extension situation without sufficient cooling of reactor core resulting core meltdown and radioactive material release even after reactor shutdown. Many safety systems had been developed like PAFS, hybrid SIT, and relocation of RPV and IRWST as a part of steps for the Fukushima accident, however, their applications have limitation in the situation that supply of feedwater into reactor is impossible due to high pressure inside reactor pressure vessel. The concept of hybrid heat pipe with control rod is introduced for breaking through the limitation. Hybrid heat pipe with control rod is the passive decay heat removal system in core, which has the abilities of reactor shutdown as control rod as well as decay heat removal as heat pipe. For evaluating the cooling performance hybrid heat pipe, a commercial CFD code, ANSYS-CFX was used. First, for validating CFD results, numerical results and experimental results with same geometry and fluid conditions were compared to a tube type heat pipe resulting in a resonable agreement between them. After that, wall temperature and thermal resistances of 2 design concepts of hybrid heat pipe were analyzed about various heat inputs. For unit length, hybrid heat pipe with a tube type of $B_4C$ pellet has a decreasing tendency of thermal resistance, on the other hand, hybrid heat pipe with an annular type $B_4C$ pellet has an increasing tendency as heat input increases.

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

• 한국압력기기공학회 논문집
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• 제12권2호
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• pp.66-70
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• 2016

Following the 2011 Fukushima Nuclear Power Plant accident, the IAEA has issued a revised version of the Nuclear Safety Standard for beyond design basis earthquake to consider the core meltdown accident. In Korea, relevant laws and regulations were also revised to consider beyond design basis earthquake to nuclear components. In this paper, CAV, an seismic damage factor that determines the restart of nuclear power plant after operating breakdown earthquake, is proposed for extension to the beyond design basis earthquake. For pipings not satisfying the beyond design basis earthquake condition, several evaluation methods are suggested, such as strain-based evaluation methods, simple nonlinear analysis method and cumulative damage evaluation method.

• Nuclear Engineering and Technology
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• 제25권2호
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• pp.215-221
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• 1993

중대 사고에 이어 형성 될 수 있는 용융 핵연료와 냉각수의 혼합체는 그 후에 일어날 수 있는 증기 폭발의 초기 조건을 이루게 되며, 또한 이러한 혼합의 정도는 증기 폭발의 강도를 결정하는 주요 요인이 된다. 본 연구는 간단한 일차원 과도 상태 모형을 이용하여 용융 핵연료와 냉각수의 혼합의 한계를 결정하였으며, 용융 핵연료의 분쇄 과정을 모사 하기 위하여 동적 분쇄모형과 순간 파열 모형을 각각 적용하여 그 결과를 비교하였다. 계산 결과에서는 용융 핵연료의 온도. 압력. 수조의 깊이, 혼합 상태에서의 분쇄물 직경 그리고 용융 핵연료 입사 직경등이 혼합량에 영향을 미치는 주요한 인자들로 나타났다.

• 원자력산업
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• 제20권6호통권208호
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• pp.31-44
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• 2000

본 연구에서 액체금속로의 노심용융(core meltdown)으로 인한 초 즉발 임계(super-prompt critical)의 출력 폭주 사고시, 노심의 반응도 및 열수력 특성 변화와 에너지 방출량등을 계산하기 위하여, Bethe-Tait 방버론을 수정, 보완한 분석 모델이 개발되었다. 주요 보완 내용으로서는, 금속 연료 노심의 단상 액체 영역에서의 선형의(Linear) threshold 형태의 상태 방정식뿐만 아니라 포화 증기(saturated fuel vapor) 영역에서의 상태 방정식이 개발되었고, 이에 따른 노심 붕괴 반응도(disassembly reactivity)의 분석 모델이 개발되었다. 또한 도플러 반응도 효과를 고려하기 위한 분석모델도 아울러 개발되었다. 상기 보완 모델을 실행할 수 있는 수치 해석 프로그램이 개발되었고, 이를 활용하여 KALIMER에서 HCDA가 발생하였을 경우 노심에서의 에너지 방출량 계산이 수행되었다. 분석결과 도플러 효과와 포화 증기 영역에서의 압력 증가 및 노심팽창의 중요성이 확인되었다. 도플러 효과가 고려되지 않을 경우 HCDA는 분석된 모든 반응도 삽입률에 대하여 폭발적인 에너지 방출과 함께 사고가 종결되는 것으로 평가되었다. 그러나 도플러 상수가 최적 평가치인 -0.002인 경우 50$/s이하의 반응도 삽입률에서는 노심은 비등점(0.8KJ/g)에 도달치 않았으며, 설계 기준 사고인 100$/s의 경우에도 노심은 포화 증기 영역에 머물고 압력이 급격히 증가하는 단상(single phase)액체 영역의 threshold 값에 미치지 않기 때문에 사고는 핵연료 증기(vapor)의 점진적인 분산과 함께 종결되는 것으로 분석되며, 총 에너지 발생량은 약 1,800MJ로서 기계적 손상 에너지로 전환되는 분율을 고려할 때 KALIMER 원자로 용기의 구조 설계 기준치에 비해 상당한 여유도를 갖는 것으로 평가되었다.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.532-548
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• 2021

Nuclear power plants contain several monitoring systems that can identify the in-vessel phenomena of a severe accident (SA). Though a lot of analysis and research is carried out on SA, right from the development of the nuclear industry, not all the possible circumstances are taken into consideration. Therefore, to improve the efficacy of the safety of nuclear power plants, additional analytical studies are needed that can directly monitor severe accident phenomena. This paper presents an interacting multiple model (IMM) based fault detection and diagnosis (FDD) approach for the identification of in-vessel phenomena to provide the accident propagation information using reactor vessel (RV) out-wall temperature distribution during severe accidents in a nuclear power plant. The estimation of wall temperature is treated as a state estimation problem where the time-varying wall temperature is estimated using IMM employing three multiple models for temperature evolution. From the estimated RV out-wall temperature and rate of temperature, the in-vessel phenomena are identified such as core meltdown, corium relocation, reactor vessel damage, reflooding, etc. We tested the proposed method with five different types of SA scenarios and the results show that the proposed method has estimated the outer wall temperature with good accuracy.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.832-845
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• 2024

Reactor cavity cooling systems (RCCSs) comprising passive safety features use the atmosphere as a coolant, which cannot be lost. However, their drawback is that they are easily affected by atmospheric disturbances. To realize the commercial application of the two types of passive RCCSs, namely RCCSs based on atmospheric radiation and atmospheric natural circulation, their safety must be evaluated, that is, they must be able to remove heat from the reactor pressure vessel (RPV) surface at all times and under any condition other than under normal operating conditions. These include both expected and unexpected natural phenomena and accidents. Moreover, they must be able to eliminate the heat leakage emitted from the RPV surface during normal operation. However, utilizing all of the heat emitted from the RPV surface increases the degree of waste heat utilization. This study aims to understand the characteristics and degree of passive safety features for heat removal by comparing RCCSs based on atmospheric radiation and atmospheric natural circulation under the same conditions. It was concluded that the proposed RCCS based on atmospheric radiation has an advantage in that the temperature of the RPV could be stably maintained against disturbances in the ambient air.

• 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.