• Nuclear Engineering and Technology
• /
• 제55권1호
• /
• 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.

• Nuclear Engineering and Technology
• /
• 제41권5호
• /
• pp.561-574
• /
• 2009

First ex-vessel core catcher has been applied to the practical design of NPPs with VVER-1000 reactors built in China (Tyanvan) and India (Kudankulam) for severe accident management (SAM) and mitigation of SA consequences. The paper presents the concept and basic design of this crucible-type core catcher as well as an evaluation of its efficiency. The important role of oxidic sacrificial material is discussed. Insight into the behaviour of the molten pool, which forms in the catcher after core relocation from the reactor vessel, is provided. It is shown that heat loads on the water-cooled vessel walls are kept within acceptable limits and that the necessary margins for departure from nucleate boiling (DNB) and of vessel failure caused by thermo-mechanical stress are satisfactorily provided for.

• 에너지공학
• /
• 제22권1호
• /
• pp.8-16
• /
• 2013

A core catcher is considered as a promising engineered system to stabilize the molten corium in the containment during a postulated severe accident in a nuclear power plant. Conceptually, the core catcher consists of a carbon steel body, sacrificial material, protection material, and engineered cooling channel. The cooling capacity of the engineered cooling channel should be guaranteed to remove the decay heat of the molten corium. The flow in ex-vessel core catcher is a combined problem of a two-phase flow in the engineered cooling channel and a single-phase natural circulation in the whole core catcher system. In this study, the analysis of the test facility for the core catcher using the CUPID code, which is a three-dimensional thermal-hydraulic code for the simulation of two-phase flows, was carried out to evaluate its cooling capacity.

• 대한기계학회논문집B
• /
• 제36권10호
• /
• pp.1043-1047
• /
• 2012

유럽지역으로의 원전 수출을 위해서는 유럽의 원전 인허가요건을 충족시켜야 하며, 이에 따르면 원전의 중대사고 대처설비로 통상 Core Catcher로 불리는 노외 노심 용융물 냉각설비를 갖출 것을 권장하고 있다. 이에 따라 본 논문에서는 노심 용융물 직접냉각방식과 간접냉각방식에 대해 각각의 개념 안의 장/단점을 비교, 검토하였으며, 그 결과 직접냉각방식은 냉각효율 측면에서, 간접냉각방식은 중대사고 사고관리 측면에서 각각 우위를 보였다.

• 에너지공학
• /
• 제23권3호
• /
• pp.21-26
• /
• 2014

유럽 원전 시장 개척을 위해 개발 중인 EU-APR1400은 중대사고 대처설비로 노외 노심용융물 보유 및 냉각을 위한 소위 Core catcher라 불리는 노외 노심용융물 냉각설비를 개발 중이며, Core catcher body를 노심용융물로부터 보호하기 위하여 노심용융물의 물성 및 상태를 변화시켜 냉각 및 보유에 유리하게 하는 희생물질을 설치한다. 중대사고 시 원자로 압력용기의 틈으로부터 노심용융물이 분출되어 희생물질에 충돌 시 열 전달량이 매우 증가하게 되므로, 이 때 노심용융물 제트의 충돌에 의한 희생물질의 침식율을 정확하게 예측하는 것은 매우 중요하다. 이 논문에서는 경계층 이론을 기반으로 한 희생물질 침식 모형을 제안하고 KAERI에서 수행한 실험결과와 비교하였다.

• 대한기계학회논문집 C: 기술과 교육
• /
• 제3권3호
• /
• pp.209-215
• /
• 2015

노내계측계통의 설치 위치 및 케이블의 관통위치가 중대사고 대처계통에 미치는 영향을 노내 노심용융물 억류 및 원자로용기 외벽냉각 전략과 노외 노심용융물 냉각계통을 중심으로 조사하였다. 기존에 국내원전에서 주로 사용되었던 노내계측계통의 원자로 용기 하부탑재 및 ICI케이블의 원자로 용기하부 관통이 중대사고에 미치는 영향을 정리하고, 이러한 단점을 개선하기 위해 노내계측계통의 ICI 케이블이 원자로 용기 상부를 관통하는 상부탑재 노내계측계통의 장점을 기술하였다.

• 대한기계학회논문집B
• /
• 제35권4호
• /
• pp.425-429
• /
• 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
• /
• 제49권7호
• /
• pp.1547-1554
• /
• 2017

VESTA (Verification of Ex-vessel corium STAbilization) and VESTA-S (-small) test facilities were constructed at the Korea Atomic Energy Research Institute in 2010 to perform various corium melt experiments. Since then, several tests have been performed for the verification of an ex-vessel core catcher design for the EU-APR1400. Ablation tests of an impinging $ZrO_2$ melt jet on a sacrificial material were performed to investigate the ablation characteristics. $ZrO_2$ melt in an amount of 65-70 kg was discharged onto a sacrificial material through a well-designed nozzle, after which the ablation depths were measured. Interaction tests between the metallic melt and sacrificial material were performed to investigate the interaction kinetics of the sacrificial material. Two types of melt were used: one is a metallic corium melt with Fe 46%, U 31%, Zr 16%, and Cr 7% (maximum possible content of U and Zr for C-40), and the other is a stainless steel (SUS304) melt. Metallic melt in an amount of 1.5-2.0 kg was delivered onto the sacrificial material, and the ablation depths were measured. Penetration tube failure tests were performed for an APR1400 equipped with 61 in-core instrumentation penetration nozzles and extended tubes at the reactor lower vessel. $ZrO_2$ melt was generated in a melting crucible and delivered down into an interaction crucible where the test specimen is installed. To evaluate the tube ejection mechanism, temperature distributions of the reactor bottom head and in-core instrumentation penetration were measured by a series of thermocouples embedded along the specimen. In addition, lower vessel failure tests for the Fukushima Daiichi nuclear power plant are being performed. As a first step, the configuration of the molten core in the plant was investigated by a melting and solidification experiment. Approximately 5 kg of a mixture, whose composition in terms of weight is $UO_2$ 60%, Zr 10%, $ZrO_2$ 15%, SUS304 14%, and $B_4C$ 1%, was melted in a cold crucible using an induction heating technique.