• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.12 no.2
• /
• pp.66-70
• /
• 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
• /
• v.25 no.2
• /
• pp.215-221
• /
• 1993

A required initial condition for a steam explosion to occur following core meltdown accidents of a nuclear power plant is the formation of a coarse mixture of molten fuel and water. The extent of a premixing is the measure of efficiency of steam explosion that may follow. A simple one-dimensional, transient model and the flooding criteria have been applied to evaluate the fuel/coolant mixing limit. Also, both instant breakup and dynamic breakup models for the mixing process have been separately used here and compared each other. The results indicate that fuel temperature, ambient pressure, mixing diameter, water depth, and pouring diameter are the important parameters affecting the mixing behavior.

• Nuclear industry
• /
• v.20 no.6 s.208
• /
• pp.31-44
• /
• 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
• /
• v.48 no.1
• /
• pp.98-113
• /
• 2016

A reactor vessel level monitoring system measures the water level in a reactor during normal operation and abnormal conditions. A drop in the water level can expose nuclear fuel, which may lead to fuel meltdown and radiation spread in accident conditions. A level monitoring system mainly consists of a sensing line and pressure transmitter. Over a period of time boron sediments or other impurities can clog the line which may degrade the accuracy of the monitoring system. The aim of this study is to determine blockage in a sensing line using the energy of the composite signal. An equivalent Pi circuit model is used to simulate blockages in the sensing line and the system's response is examined under different blockage levels. Composite signals obtained from the model and plant's unblocked and blocked channels are decomposed into six levels of details and approximations using a wavelet filter bank. The percentage of energy is calculated at each level for approximations. It is observed that the percentage of energy reduces as the blockage level in the sensing line increases. The results of the model and operational data are well correlated. Thus, in our opinion variation in the energy levels of approximations can be used as an index to determine the presence and degree of blockage in a sensing line.

• Nuclear Engineering and Technology
• /
• v.48 no.2
• /
• pp.448-456
• /
• 2016

Severe accident scenarios in nuclear reactors, such as nuclear meltdown, reveal that an extremely hot molten core may fall into the nuclear reactor cavity and seriously affect the safety of the nuclear containment vessel due to the chain reaction caused by the reaction between the molten core and concrete. This paper reports on research focused on the type and amount of vapor produced during the reaction between a high-temperature molten core and concrete, as well as on the erosion rate of concrete and the heat transfer characteristics at its vicinity. This study identifies themass fraction and melting temperature as the most influential properties of concrete necessary for a safety analysis conducted in relation to the thermal interaction between the molten core and the basemat concrete. The types of concrete that are actually used in nuclear reactor cavities were investigated. The $H_2O$ content in concrete required for the computation of the relative amount of gases generated by the chemical reaction of the vapor, the quantity of $CO_2$ necessary for computing the cooling speed of the molten core, and the melting temperature of concrete are evaluated experimentally for the molten core-concrete interaction analysis.

• Nuclear Engineering and Technology
• /
• v.53 no.2
• /
• pp.532-548
• /
• 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
• /
• v.55 no.1
• /
• pp.144-155
• /
• 2023

The core catcher is used as a passive safety system in new generation nuclear power plants to create a space in the containment for the placing and cooling of the molten corium under various severe accidents. This research investigates the role of the core catcher in the VVER-1000 reactor containment system in mitigating the effects of core meltdown under various severe accidents within the context of the Ex-vessel Melt Retention (EVMR) strategy. Hence, a comparison study of three severe accidents is conducted, including Station Black-Out (SBO), SBO combined with the Large Break Loss of Coolant Accident (LB-LOCA), and SBO combined with the Small Break Loss of Coolant Accident (SB-LOCA). Numerical comparative simulations are performed for the aforementioned scenario with and without the EX-vessel core-catcher. The results showed that considering the EX-Vessel core catcher reduces the amount of hydrogen by about 18.2 percent in the case of SBO + LB-LOCA, and hydrogen production decreases by 12.4 percent in the case of SBO + SB-LOCA. Furthermore, in the presence of an EX-Vessel core-catcher, the production of gases such as CO and CO2 for the SBO accident is negligible. It was revealed that the greatest decrease in pressure and temperature of the containment is related to the SBO accident.

• Proceedings of the Korea Water Resources Association Conference
• /
• 2022.05a
• /
• pp.275-275
• /
• 2022

Tsunamis triggered by undersea earthquakes have the characteristic of longer wavelengths and can propagate a very long distance. Although the occurrence frequency of tsunami is low, it can cause casualties and properties. Historically, tsunamis that occurred on the western coast of Japan attacked the eastern coast of the Korean Peninsula and damaged the property and the loss of human life in 1983 and 1993. By tsunami in 1983 especially, 2 people were killed, and more than 200 casualties occurred. In addition, it caused 2 million dollars in property damage at Imwon Port. In 2011, The eastern cities of Japan: Iwate, Miyagi, Ibaraki, and Fukushima were damaged by a tsunami that occurred near onshore along the Pacific ocean and caused more than 300 billion dollars in property damage, and 20,000 casualties occurred. Moreover, those provoked nuclear power plant meltdown at Fukushima. In this study, it was carried out a relationship between maximum tsunami heights and fault parameters of earthquake: strike angle, dip angle, and slip angle at Imwon port. Those fault parameters are known that it does not relate to the magnitude of earthquake directly. Virtual tsunamis, which could be triggered by probable undersea earthquakes in the future, were investigated and mutual information based on probability and information theory was introduced to figure out the relationship between maximum tsunami height and fault parameters. Fault parameters were evaluated according to the strong relationship with maximum tsunami heights finally.

• Nuclear Engineering and Technology
• /
• v.56 no.3
• /
• pp.832-845
• /
• 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.

• Nuclear Engineering and Technology
• /
• v.49 no.7
• /
• pp.1442-1450
• /
• 2017

In liquid metal fast breeder reactors, postulated failures of the plant protection system may lead to serious unprotected accidental consequences. Unprotected transients are generically categorized as transient overpower accidents and transient under cooling accidents. In both cases, core meltdown may occur and this can lead to a molten fuel coolant interaction (MFCI). The understanding of MFCI phenomena is essential for study of debris coolability and characteristics during post-accident heat removal. Sodium is used as coolant in liquid metal fast breeder reactors. Viewing inside sodium at elevated temperature is impossible because of its opaqueness. In the present study, a methodology to depict MFCI phenomena using a flat panel detector based imaging system (i.e., real time radiography) is brought out using a woods metal-water experimental facility which simulates the $UO_2-Na$ interaction. The developed imaging system can capture attributes of the MFCI process like jet breakup length, jet front velocity, fragmented particle size, and a profile of the debris bed using digital image processing methods like image filtering, segmentation, and edge detection. This paper describes the MFCI process and developed imaging methodology to capture MFCI attributes which are directly related to the safe aspects of a sodium fast reactor.