• Nuclear Engineering and Technology
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• 제54권3호
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• pp.896-908
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• 2022

As a part of probabilistic risk (or safety) assessment (PRA or PSA) of nuclear power plants (NPPs), the primary role of human reliability analysis (HRA) is to provide credible estimations of the human error probabilities (HEPs) of safety-critical tasks. In this regard, it is vital to provide credible HEPs based on firm technical underpinnings including (but not limited to): (1) how to collect HRA data from available sources of information, and (2) how to inform HRA practitioners with the collected HRA data. Because of these necessities, the U.S. Nuclear Regulatory Commission and the Korea Atomic Energy Research Institute independently developed two dedicated HRA data collection systems, SACADA (Scenario Authoring, Characterization, And Debriefing Application) and HuREX (Human Reliability data EXtraction), respectively. These systems provide unique frameworks that can be used to secure HRA data from full-scope training simulators of NPPs (i.e., simulator data). In order to investigate the applicability of these two systems, two papers have been prepared with distinct purposes. The first paper, entitled "SACADA and HuREX: Part 1. The Use of SACADA and HuREX Systems to Collect Human Reliability Data", deals with technical issues pertaining to the collection of HRA data. This second paper explains how the two systems are able to inform HRA practitioners. To this end, the process of estimating HEPs is demonstrated based on feed-and-bleed operations using HRA data from the two systems.

• Nuclear Engineering and Technology
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• 제56권2호
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• pp.762-769
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• 2024

The TMSR-SF0 simulator is an integral effect thermal-hydraulic experimental system for the development of thorium molten salt reactor (TMSR) program in China. The simulator has two heat transport loops with liquid FLiNaK. In literature, the 95% level confidence uncertainties of the thermophysical properties of FLiNaK are recommended, and the uncertainties of density, heat capacity, thermal conductivity and viscosity are ±2%, ±10, ±10% and ±10% respectively. In order to investigate the effects of thermophysical properties uncertainties on the molten salt heat transport system, the uncertainty and sensitivity analysis of the heat transfer characteristics of the simulator system are carried out on a RELAP5 model. The uncertainties of thermophysical properties are incorporated in simulation model and the Monte Carlo sampling method is used to propagate the input uncertainties through the model. The simulation results indicate that the uncertainty propagated to core outlet temperature is about ±10 ℃ with a confidence level of 95% in a steady-state operation condition. The result should be noted in the design, operation and code validation of molten salt reactor. In addition, more experimental data is necessary for quantifying the uncertainty of thermophysical properties of molten salts.

• 한국의학물리학회지:의학물리
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• 제20권2호
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• pp.51-61
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• 2009

컴프턴 카메라는 컴프턴 산란현상의 기하학적 해석을 통해 감마선원의 3차원적 위치분포를 고정된 위치에서 찾아내는 신개념의 감마선 영상장치이다. 기존의 감마선 영상장치에서는 필수적으로 사용되는 기계적 집속기를 사용하지 않기 때문에 높은 영상 감도를 제공할 수 있으며, 다중 추적자 기능을 제공한다는 장점이 있어 차세대 감마선 영상장치로서 주목을 받고 있다. 본 연구에서는 Geant4 몬테칼로 전산모사 툴키트를 이용하여 사용자 친화형의 컴프턴 카메라 시뮬레이터를 개발하였다. 시뮬레이터의 정확성을 검증하기 위하여 한양대학교에서 개발 중인 이중 산란형 컴프턴 카메라의 실험 결과와 비교하였다. 영상 해상도의 경우 선원의 에너지가 높아짐에 따라 해상도가 향상되는 동일한 경향을 보였으나, 시뮬레이터 결과에서 약 1 mm 정도 우수하게 평가되었다. 영상 감도의 경우 실험에서 2~3배 정도 높게 평가되었다. 이러한 결과들은 우연 동시반응에 기인한 것으로, 실험에서 획득한 유효반응들은 전산모사와는 달리 우연한 동시계수에의한 데이터가 상당수 포함되어 있기 때문에 영상 해상도는 상대적으로 저조하고, 영상 감도는 높게 나타난 것으로 판단된다. 개발된 시뮬레이터를 이용하여 새로운 구조의 컴프턴 카메라에 대한 성능을 평가하였다. 단일 산란형 컴프턴 카메라의 경우 산란부를 다층 구조로 개발할 때 4대의 산란부 검출기를 사용하는 것이 최상의 성능을 보이는 것으로 평가되었고, 그 이상 사용할 경우 성능이 오히려 저하됨을 확인하였다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.219-226
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• 1997

This paper presents the development of an Integrated Test Facility (ITF) that consists of a full-scope Human Machine Simulator (HMS) and some experimental measurement systems. This facility is aimed to experiment with tile design of the control room environment and the human interaction as it relates to the control of advanced nuclear power plants.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.117-122
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• 1995

An Integrated Test Facility(ITF) is a human factors experimental environment to evaluate an advanced man machine interface(MMI) design. The ITF includes a human machine simulator(HMS) comprised of a nuclear power plant function simulator, man-machine interface, experiment control station for the experiment control and design, human behavioural data measurement system, and data analysis and experiment evaluation supporting system(DAEXESS). The most important features of ITF is to secure the flexibility and expandibility of Man Machine Interlace(MMI) design to change easily the environment of experiments to accomplish the experiment's objects In this paper, we describe a development scope and characteristics of the ITF such as, hardware and software development scope and characteristics, system thermohydraulic modelling characteristics, and experiment station characteristics for the experiment variables design and control, to be used as an experiment environment for the evaluation of VDU-based control room.

• 한국정보처리학회:학술대회논문집
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• 한국정보처리학회 2011년도 춘계학술발표대회
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• pp.547-550
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• 2011

The purpose of this paper is to verify and validate the maintenance tasks of the construction of a nuclear facility using a digital mock-up and simulation technology instead of a physical mock-up. Prior to the construction of a nuclear facility, a remote simulator that provides the opportunity to produce a complete digital mock-up of the PRIDE (Pyroprocess Integrated Inactive DEmonstration Facility) region and its remote handling equipment, including operations and maintenance procedures has been developed. In this paper, the system architecture and graphic user interface of a remote simulator that coincides with the extraordinary nature of a nuclear fuel cycle facility is introduced. In order to analyze the remote accessibility of a remote manipulator, virtual prototyping that was performed it by using haptic device of external input devices under a 3D full-scale digital mock-up is explained.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 추계학술발표회요약집
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• pp.390-390
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• 2001

• Nuclear Engineering and Technology
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• 제47권2호
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• pp.176-186
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• 2015

Pattern classifications have become important tools for fault diagnosis in nuclear power plants (NPP). However, it is often difficult to obtain training data under fault conditions to train a supervised classification model. By contrast, normal plant operating data can be easily made available through increased deployment of supervisory, control, and data acquisition systems. Such data can also be used to train classification models to improve the performance of fault diagnosis scheme. In this paper, a fault diagnosis scheme based on semisupervised classification (SSC) scheme is developed. In this scheme, new measurements collected from the plant are integrated with data observed under fault conditions to train the SSC models. The trained models are subsequently applied to new measurements for fault diagnosis. In comparison with supervised classifiers, the proposed scheme requires significantly fewer data collected under fault conditions to train the classifier. The developed scheme has been validated using different fault scenarios on a desktop NPP simulator as well as on a physical NPP simulator using a graph-based SSC algorithm. All the considered faults have been successfully diagnosed. The results have demonstrated that SSC is a promising tool for fault diagnosis in NPPs.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2016년도 추계학술논문요약집
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• pp.351-352
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• 2016

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.385-386
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• 2004