• Nuclear Engineering and Technology
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• 제17권1호
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• pp.53-67
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• 1985

중대 노심사고는 아직 Design Basis Accident에 포함되지 않고 있으나, 극히 적은 사고 확률을 가지는 반면 사고 후 영향이 큼으로해서 원자력발전소의 전반적 위험 평가에 중요한 요인중의 하나가 되고 있다. 중대 노심사고시 격납용기 손상에 관련된 물리현상들은 Steam Explosion, Debris Bed Coolability, Hydrogen Burning, Steam Spike와 Core-Concrete Interaction 등이며, 각각의 현상에 대한 좀 더 나은 이해를 위해 현재 이루어지고 있는 연구들에 대한 개략적 설명을 시도 하였다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.61-62
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• 2005

• Nuclear Engineering and Technology
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• 제51권8호
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• pp.1916-1938
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• 2019

The objective of this paper is to discuss the modeling principles of phenomena governing core degradation/melting and in-vessel melt relocation during severe accidents in light water reactors. The proposed modeling approach has been applied in the development of a new accident simulation package, COMPASS (COre Meltdown Progression Accident Simulation Software). COMPASS can be used either as a stand-alone tool to simulate in-vessel meltdown progression up to and including RPV failure, or as a component of an integrated simulation package being developed in Korea for the APR1400 reactor. Interestingly, since the emphasis in the development of COMPASS modeling framework has been on capturing generic mechanistic aspects of accident progression in light water reactors, several parts of the overall model should be useful for future accident studies of other reactor designs, both PWRs and BWRs. The issues discussed in the paper include the overall structure of the model, the rationale behind the formulation of the governing equations and the associated simplifying assumptions, as well as the methodology used to verify both the physical and numerical consistencies of the overall solver. Furthermore, the results of COMPASS validation against two experimental data sets (CORA and PHEBUS) are shown, as well as of the predicted accident progression at TMI-2 reactor.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.144-155
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• 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
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• 제34권2호
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• pp.117-131
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• 2002

Core disruptive accidents have been investigated at Korea Atomic Energy Research Institute(KAERI) as part of the work to demonstrate the inherent and ultimate safety of conceptual design of the Korea Advanced Liquid Metal Reactor(KALIMER), a 150 MWe pool- type sodium cooled prototype fast reactor that uses U-Pu-Zr metallic fuel. In this study, a simple method and associated computer program, SCHAMBETA, was developed using a modified Bethe-Tait method to simulate the kinetics and thermodynamic behavior of a homogeneous spherical core over the period of the super-prompt critical power excursion induced by the ramp reactivity insertion. Calculations of the energy release during excursions in the sodium-voided core of the KALIMER were subsequently performed using the SCHAMBETA code for various reactivity insertion rates up to 100 S/s, which has been widely considered to be the upper limit of ramp rates due to fuel compaction. Benchmark calculations were made to compare with the results of more detailed analysis for core meltdown energetics of the oxide fuelled fast reactor. A set of parametric studies were also performed to investigate the sensitivity of the results on the various thermodynamics and reactor parameters.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.3990-4002
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• 2021

CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis.

• 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.

• 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.

• 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.