• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2670-2677
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• 2023

In the nuclear safety studies, a new trend refers to the evaluation of uncertainties as a mandatory component of best-estimate safety analysis which is a modern and technically consistent approach being known as BEPU (Best Estimate Plus Uncertainty). The major objectives of this study consist in performing a study of uncertainties/sensitivities of the major analysis results for a generic CANDU 6 Nuclear Power Plant during Station Blackout (SBO) progression to understand and characterize the sources of uncertainties and their effects on the key figure-of-merits (FOMs) predictions in severe accidents (SA). The FOMs of interest are hydrogen mass generation and event timings such as the first fuel channel failure time, beginning of the core disassembly time, core collapse time and calandria vessel failure time. The outcomes of the study, will allow an improvement of capabilities and expertise to perform uncertainty and sensitivity analysis with severe accident codes for CANDU 6 Nuclear Power Plant.

• 에너지공학
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• 제26권3호
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• pp.64-70
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• 2017

한국원자력연구원에서 제안한 혼합형 안전주입탱크 (Hybrid SIT)는 APR+ 원자로에 적용하기 위해 개발된 피동안전주입시스템이다. 본 연구는 대표적인 고압사고인 발전소정전사고 시 Hybrid SIT의 냉각성능을 평가하기 위해 열수력 안전해석 코드인 MARS-KS 코드를 이용한 예비해석에 대한 것이다. PAFS 구동이 정지되면, 열제거량이 감소하게 되어 가압기와 증기발생기의 압력이 상승하기 시작하며, 가압기의 압력이 안전감압계통(Pilot Operated Safety and Relief Valve) 개방 설정치인 17.03 MPa에 도달하면, 그와 동시에 Hybrid SIT의 증기격리밸브가 열림으로서 가압기 상단의 증기가 Hybrid SIT로 주입되게 된다. 주입된 증기에 의해 압력평형이 빠른시간 안에 이루어졌으며, 주입배관을 통해 냉각수가 주입 되었다. 발전소정전사고시 PAFS와 같은 열제거수단이 상실됨에도 혼합형 Hybrid SIT가 주입되는 시간동안은 노심의 수위가 유지됨을 확인할 수 있었고, 수위가 유지됨에 따라 노심 출구 온도(CET)의 상승을 방지함을 확인하였다.

• 한국해양공학회지
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• 제36권5호
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• pp.340-352
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• 2022

This article presents a modeling method for the uncorrelated measurement error of the ultra-short baseline (USBL) acoustic positioning system for aiding navigation of underwater vehicles. The Mahalanobis distance (MD) and principal component analysis are applied to decorrelate the errors of USBL measurements, which are correlated in the x- and y-directions and vary according to the relative direction and distance between a reference station and the underwater vehicles. The proposed method can decouple the radial-direction error and angular direction error from each USBL measurement, where the former and latter are independent and dependent, respectively, of the distance between the reference station and the vehicle. With the decorrelation of the USBL errors along the trajectory of the vehicles in every time step, the proposed method can reduce the threshold of the outlier decision level. To demonstrate the effectiveness of the proposed method, simulation studies were performed with motion data obtained from a field experiment involving an autonomous underwater vehicle and USBL signals generated numerically by matching the specifications of a specific USBL with the data of a global positioning system. The simulations indicated that the navigation system is more robust in rejecting outliers of the USBL measurements than conventional ones. In addition, it was shown that the erroneous estimation of the navigation system after a long USBL blackout can converge to the true states using the MD of the USBL measurements. The navigation systems using the uncorrelated error model of the USBL, therefore, can effectively eliminate USBL outliers without loss of uncontaminated signals.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1414-1422
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• 2017

Remote response technology has advanced to the extent that a robot system, if properly designed and deployed, may greatly help respond to beyond-design-basis accidents at nuclear power plants. Particularly in the aftermath of the Fukushima accident, there is increasing interest in developing disaster robots that can be deployed in lieu of a human operator to the field to perform mitigating actions in the harsh environment caused by extreme natural hazards. The nuclear robotics team of the Korea Atomic Energy Research Institute (KAERI) is also endeavoring to construct disaster robots and, first of all, is interested in finding out to what extent safety benefits can be achieved by such a disaster robotic system. This paper discusses a new approach based on the probabilistic risk assessment (PRA) technique, which can be used to quantify safety benefits associated with disaster robots, along with a case study for seismic-induced station blackout condition. The results indicate that to avoid core damage in this special case a robot system with reliability > 0.65 is needed because otherwise core damage is inevitable. Therefore, considerable efforts are needed to improve the reliability of disaster robots, because without assurance of high reliability, remote response techniques will not be practically used.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.587-591
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• 1997

The DGs (Diesel Generators) in NPP (Nuclear Power Plant) has been used for the emergency electric power source to shut down the nuclear reactor safely in case of station blackout. The RCM (Reliability Centered Maintenance) has been applied to DGs for increasing the safety of NPP. The structured defects of DG were not remedied by the improvement of maintenance method. As the first stage of RCM, to find the structured defects, its failure modes were searched and analyzed through the ten year maintenance information. The structured defects such as the air compressor, the lubricating oil pressure, and the insufficient load were the root causes of main failures. The air reservoir reinstallation, the lubricating oil tube modification, the load bank installation, and the qualitative instrumentation were the solutions for the hardware oriented RCM of DGs. There remains the software oriented RCM such as the rejection of useless maintenance, the preventive maintenance, the database of maintenance information, and the predictive maintenance.

• 한국안전학회지
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• 제17권3호
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• pp.123-130
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• 2002

본 논문은 사고관리방안 수행에 있어서 발생되는 인적오류의 정량적 평가방법을 개발하여 공동범람 사고관리방안의 예제문제에 적용한 연구결과를 기술하고있다. PSA에서 사용되었던 기존의 인간오류평가 방법론인 THERP, HCR, SLIM-MAUD 방법의 특징을 검토하여 장단점을 기술하였다. 본 연구에서 제시하는 인간오류평가 방법론은 신뢰도물리모델을 이용하는 새로운 HRA 분석방법이다. 불확실성 분석을 위하여 MAAP 코드와 LHS 코드가 사용되었다. 본 연구를 통하여 제안하는 방법은 매우 유연하여 중대사고관리방안과 관련한 다양한 인간오류행위에 대한 평가에 사용될 수 있음을 보여주었다.

• 시스템엔지니어링학술지
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• 제12권2호
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• pp.91-99
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• 2016

The safety of nuclear power plants has received much attention; this safety largely depends on the continuous availability of electrical energy source during all modes of nuclear power plant operation. A station blackout (SBO) describes the loss of the off-site electric power, the failure of the emergency diesel generators, and the unavailability of the alternate AC (AAC) power. Consequently, all systems that are AC powered such as the safety injection, shutdown cooling, component cooling water, and essential service water systems are unavailable. The aim of this study is to investigate the deficiencies of the existing alternatives for coping with an extended SBO for APR1400 design. The method is analyzing the existing deficiencies and proposing an optimal solution for the NPP design during the extended SBO. This study, established a new passive system, called passive decay heat removal system (PDHRS), using systems engineering approach.

• 전자공학회논문지
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• 제51권3호
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• pp.10-21
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• 2014

후쿠시마 원전사고는 자연재해에 의한 중대사고 발생 시 전원공급 중단 및 극한 환경으로 인해 발전소 내부 상황을 정확하게 파악하지 못하였고, 대부분의 계측제어시스템이 그 기능을 제대로 발휘하지 못해 비상냉각기능이 상실되어 수소폭발 및 다량의 방사능이 누출된 사고였다. 본 논문에서는 중대사고 발생 시에도 발전소 내부 상황을 감시하고, 적절히 제어할 수 있는 비상대응시스템에 대하여 소개하고, 비상대응시스템에 사용되는 무선통신망의 성능요구사항에 대해서 논의하고, 요구사항을 만족시킬 수 있는 비상통신망의 성능을 마코프 모델을 이용하여 분석하였다.

• Nuclear Engineering and Technology
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• 제49권2호
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• pp.360-372
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• 2017

Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event.

• Nuclear Engineering and Technology
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• 제47권1호
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• pp.66-73
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• 2015

Background: A numerical simulation was carried out to investigate the difference between internal and external heat-flux distributions at the reactor vessel wall under in-vessel retention through external reactor vessel cooling (IVR-ERVC). Methods: Total loss of feed water, station blackout, and large break loss of coolant accidents were selected as the severe accident scenarios, and a transient analysis using the element-birth-and-death technique was conducted to reflect the vessel erosion (vessel wall thickness change) effect. Results: It was found that the maximum heat flux at the focusing region was decreased at least 10% when considering the two-dimensional heat conduction at the reactor vessel wall. Conclusion: The results show that a higher thermal margin for the IVR-ERVC strategy can be achieved in the focusing region. In addition, sensitivity studies revealed that the heat flux and reactor vessel thickness are dominantly affected by the molten corium pool formation according to the accident scenario.