• Nuclear Engineering and Technology
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• 제40권5호
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• pp.397-408
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• 2008

This paper describes the application of a software fault tree analysis (FTA) as one of the analysis techniques for a software safety analysis (SSA) at the design phase and its analysis results for the safety-critical software of a digital reactor protection system, which is called the KNICS RPS, being developed in the KNICS (Korea Nuclear Instrumentation & Control Systems) project. The software modules in the design description were represented by function blocks (FBs), and the software FTA was performed based on the well-defined fault tree templates for the FBs. The SSA, which is part of the verification and validation (V&V) activities, was activated at each phase of the software lifecycle for the KNICS RPS. At the design phase, the software HAZOP (Hazard and Operability) and the software FTA were employed in the SSA in such a way that the software HAZOP was performed first and then the software FTA was applied. The software FTA was applied to some critical modules selected from the software HAZOP analysis.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.928-940
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• 2017

An experiment using the $Prim{\ddot{a}}rkreisl{\ddot{a}}ufe$ Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg smallbreak loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.979-988
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• 2017

The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.2915-2923
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• 2022

There is an increasing interest in passive safety systems to minimize the need for operator intervention or external power sources in nuclear power plants. Because a passive system has a weak driving force, there is greater uncertainty in the performance compared with an active system. In previous studies, several methods have been suggested to evaluate passive system reliability, and many of them estimated the failure probability using thermal-hydraulic analyses and the Monte Carlo method. However, if the functional failure of a passive system is rare, it is difficult to estimate the failure probability using conventional methods owing to their high computational time. In this paper, a procedure for the application of the Chernoff bound to the evaluation of passive system reliability is proposed. A feasibility study of the procedure was conducted on a passive decay heat removal system of a micro modular reactor in its conceptual design phase, and it was demonstrated that the passive system reliability can be evaluated without performing a large number of thermal-hydraulic analyses or Monte Carlo simulations when the system has a small failure probability. Accordingly, the advantages and constraints of applying the Chernoff bound for passive system reliability evaluation are discussed in this paper.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.131-143
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• 2023

Constitutive equations in a nuclear reactor safety analysis code are mostly empirical correlations developed from experiments, which always accompany uncertainties. The accuracy of the code can be improved by modifying the constitutive equations fitting wider range of data with less uncertainty. Thus, the sensitivity of the code with respect to the constitutive equations is evaluated quantitatively in the paper to understand the room for improvement of the code. A new methodology is proposed which first starts by dividing the thermal hydraulic conditions into multiple sub-regimes using self-organizing map (SOM) clustering method. The sensitivity analysis is then conducted by multiplying an arbitrary set of coefficients to the constitutive equations for each sub-divided thermal-hydraulic regime with SOM to observe how the code accuracy varies. The randomly chosen multiplier coefficient represents the uncertainty of the constitutive equations. Furthermore, the set with the smallest error with the selected experimental data can be obtained and can provide insight which direction should the constitutive equations be modified to improve the code accuracy. The newly proposed method is applied to a steady-state experiment and a transient experiment to illustrate how the method can provide insight to the code developer.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.873-879
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• 2024

A sodium-cooled fast reactor (SFR) core has a potential of prompt criticality due to a change of core material distribution during a severe accident, and the resultant energy release has been one of the safety issues of SFRs. In this study, the safety assessment of an unprotected loss-of-flow (ULOF) in a small SFR (SSFR) has been performed using the SIMMER-IV computer code, which couples the models of space- and time-dependent neutronics and multi-component, multi-field thermal hydraulics in three dimensions. The code, therefore, is applicable to the simulations of transient behaviors of extended disrupted core material motion and its reactivity effects during the transition phase (TP) of ULOF, including a potential of prompt-criticality power excursions driven by fuel compaction. Several conservative assumptions are used in the TP analysis by SIMMER-IV. It was found out that one of the important mechanisms that drives the reactivity-inserting fuel motion was sodium vapor pressure resulted from a fuel-coolant interaction (FCI), which itself was non-energetic local phenomenon. The uncertainties relating to FCI is also evaluated in much conservative way in the sensitivity analysis. From this study, the ULOF characteristics in an SSFR have been understood. Occurrence of recriticality events under conservative assumptions are plausible, but their energy releases are limited.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4607-4616
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• 2023

We implement machine learning regression models to predict peak pressures of primary and secondary systems, a major safety concern in Loss Of Condenser Vacuum (LOCV) accident. We selected the Multi-dimensional Analysis of Reactor Safety-KINS standard (MARS-KS) code to analyze the LOCV accident, and the reference plant is the Korean Optimized Power Reactor 1000MWe (OPR1000). eXtreme Gradient Boosting (XGBoost) is selected as a machine learning tool. The MARS-KS code is used to generate LOCV accident data and the data is applied to train the machine learning model. Hyperparameter optimization is performed using a simulated annealing. The randomly generated combination of initial conditions within the operating range is put into the input of the XGBoost model to predict the peak pressure. These initial conditions that cause peak pressure with MARS-KS generate the results. After such a process, the error between the predicted value and the code output is calculated. Uncertainty about the machine learning model is also calculated to verify the model accuracy. The machine learning model presented in this paper successfully identifies a combination of initial conditions that produce a more conservative peak pressure than the values calculated with existing methodologies.

• 전산구조공학
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• 제7권3호
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• pp.95-103
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• 1994

원자력발전소 부품중 안전과 관련된 구조물은 지진하중하에서 그 건전성을 유지하도록 설계되어야 한다. 그중 원자로내부구조물부품은 1차 내진분류에 속하는 것으로써 지진하중하에서의 건전성이 발전소 안전과 경제적인 관점에서 매우 중요하다. 지금까지 이러한 원자로내부구조물의 모델링에 대해서는 여러 사람들에 의해 연구되고 발표되었으나, 본 논문에서는 국내 발전소 중에서 Turn-jey base로 건설되어 이미 가동 중에 있는 영광 1&2호기의 원자로내부구조물에 대한 안전정지지진하의 거동을 Global Beam Model이라는 단순화된 모델을 이용하여 분석하였다. 이 모델의 설정을 위해서 주요부품들을 double pendulum의 보요소로 표현하였고, 이들 주요부품들의 특성해석을 범용유한 요소해석 코드인 ANSYS에 의해 구하여 이를 상부 및 하부에서 간격을 갖는 비선형 스프링으로 모델링하였다. 또한 이 비선형 스프링뿐만아니라 원자로용기와 원자로내부구조물부품들 사이의 유체동적현상을 묘사한 유체동력학적 coupling에 의해 pendulum의 보요소를 서로 연결시켜 모델링을 하였다. 가진자료인 안전정지하중은 영광 1&2호기의 원자로용기 지지부에 가해지는 응답스펙트럼을 시간이력함수로 바꾸었으며, 이 모델과 간진 하중을 가지고 비선형해석 code인 KWUSTOSS의 explicit Runge-Kutta-Gills algorithm을 이용하여 적분을 수행하므로써 안전정지지진하의 원자로 내부구조물에 대한 거동을 구하여 이 구조물의 주요부품에 대한 내진검증 및 구조물 내부에 있는 핵연료집합체의 내진 해석을 위한 입력자료를 확보할 수 있었다. 그리고 본 연구에서 사용된 Globa Beam Model의 간편성 및 효율성과 explicit Runge-Kutta-Gills algorithm에 대한 경제성을 확인할 수 있었다.파악되었 다. 그 외에도 '옥외공간이용 편리'(outdoor or recreation convenience)와 ' 이웃만족'(satisfaction with neighbors), 그리고 '주거환경 유형'(building type, building arrangement type)등도 유의한 인과적 관련을 보이므로써, 기존 문헌들이 제시하고 있는 것보다 훨씬 다양한 변수들이 다양한 경로를 통해 거주자 시각만족의 영향인자가 될 수 있는 가능성을 제시하고 있다. 가설 변수의 하나인 '길찾기의 난이 정도'(difficulty of way-finding)와 종 속변수간에 유의한 관련도가 나타나지 않은 이유로 길찾기 변수가 '시각만 족'보다는 거주자의 '안전만족'(safety)과 관련된 변수일 가능성도 아울러 지적되었다. 본 연구의 결과로부터, 주거 계획 및 설계분야 그리고 추후 관 련 연구 분야를 위한 여러 제안들이 제시되었다.에 관한 국가 규격은 국제 규격에서 저술한 바와 같이 특별히 규정된 것이 없고 VDE(Verband Deutscher Elektrotechniker: 서독전기기술 협회)와 SAE(Society of Automotive Engi- neers: 자동차 기술자 협회)에서 비교적 활발하고 Jaso(Japanese Automobile Standards Organization: 일본 자동차 표준협회)에서 많이 진행중에 있다. 본 고에서는 자동차의 전자제어에 따른 잡음 발생 요인과 전자파 간섭 관련 자동차 규격과 시험평가 방법에 대해 간단히 소개 하였다.저하에 저해요인으로서가 아니라, 인위적이던 자연적이던 간에 아들만 두면 단산하는 현행의 출산풍토하에서는 남아선호관이 오히려 출산력저하에 결정적으로 작용하고 있다고 하겠다. 태아의 성 판별을 통한 선택적 인공임신중절의 건수는 1990년 한해에

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집D
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• pp.921-926
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• 2001

The reliability and accuracy of the information on control rod position are very important to the reactor safety and the design of the core protection system. In this study, a thorough investigation on the RSPT(Reed Switch Position Transmitter) type control rod position indication system and its actual implementation in the exiting nuclear power plants in Korea was performed first. A design of the control rod position indication system using reed switch for the CEDM on the system-integrated reactor SMART was developed based on the position indicator technology identified through the investigation. The feasibility of the design was evaluated by test of manufactured control rod position indicator using reed switch for SMART.

• 한국건축시공학회:학술대회논문집
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• 한국건축시공학회 2012년도 추계 학술논문 발표대회
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• pp.167-168
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• 2012

Analysis on structural effects which are reduction of friction coefficient and increase of tendon area by HDPE greased and large size strand in post-tensioning system of reactor containment building was carried out. Effective ratio of tendon force increases 67% to 83% by HDPE greased strand and vertical, horizontal internal section forces increased maximum 51%, 41% respectively. Tendon quantity could be reduced 30% by large size and HDPE greased strand that can maintain safety of ultimate internal pressure same as at present.