• Nuclear Engineering and Technology
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• 제47권2호
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• pp.204-211
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• 2015

Probabilistic safety assessment (PSA) has had a significant role in quantitative decision-making by finding design and operational vulnerabilities and evaluating cost-benefit in improving such weak points. In particular, it has been widely used as the core methodology for risk-informed applications (RIAs). Even though the nature of PSA seeks realistic results, there are still "conservative" aspects. One of the sources for the conservatism is the assumptions of safety analysis and the estimation of failure frequency. Surveillance, diagnosis, and prognosis (SDP), utilizing massive databases and information technology, is worth highlighting in terms of its capability for alleviating the conservatism in conventional PSA. This article provides enabling techniques to solidify a method to provide time- and condition-dependent risks by integrating a conventional PSA model with condition monitoring and prognostics techniques. We will discuss how to integrate the results with frequency of initiating events (IEs) and probability of basic events (BEs). Two illustrative examples will be introduced: (1) how the failure probability of a passive system can be evaluated under different plant conditions and (2) how the IE frequency for a steam generator tube rupture (SGTR) can be updated in terms of operating time. We expect that the proposed model can take a role of annunciator to show the variation of core damage frequency (CDF) depending on operational conditions.

• Nuclear Engineering and Technology
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• 제42권4호
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• pp.450-459
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• 2010

The present paper investigates the collapse pressure of cylinders with intermediate thickness subjected to external pressure based on detailed elastic-plastic finite element (FE) analyses. The effect of the initial ovality of the tube on the collapse pressure was explicitly considered in the FE analyses. Based on the present FE results, the analytical yield locus, considering the interaction between the plastic collapse and local instability due to initial ovality, was also proposed. The collapse pressure values based on the proposed yield locus agree well with the present FE results; thus, the validity of the proposed yield locus for the thickness range of interest was verified. Moreover, the partial safety factor concept based on the structural reliability theory was also applied to the proposed collapse pressure estimation model, and, thus, the priority of importance of respective parameter constituting for the collapse of cylinders under external pressure was estimated in this study. From the application of the partial safety factor concept, the yield strength was concluded to be the most sensitive, and the initial ovality of tube was not so effective in the proposed collapse pressure estimation model. The present deterministic and probabilistic results are expected to be utilized in the design and maintenance of cylinders subjected to external pressure with initial ovality, such as the once-through type steam generator.

• 한국신재생에너지학회:학술대회논문집
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• 한국신재생에너지학회 2009년도 춘계학술대회 논문집
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• pp.323-327
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• 2009

To extract hydrogen for stack, fuels such as LPG and LNG were reformed in the fuel processor, which is comprised of desulfurizer, reformer, shift converter, CO remover and steam generator. All elements of fuel processor are integrated in a single package. Highly active catalysts (desulfurizing adsorbent, reforming catalyst, CO shift catalyst, CO removal catalyst) and the various burners were developed and evaluated in this study. The performance of the developed catalysts and the commercial ones was similar. 1 kW, 5 kW class fuel processor systems using the developed catalyst and burner showed efficiency of 75 %(LHV, for LNG). The start-up time of the 1 kW class fuel processor was less than 50 minutes and its volume including insulation was about 30 l. The start-up time of 3 kW and 5 kW class fuel processors with the volume of 90 l and 150 l, respectively, was about 60 minutes. In the case of LPG fuel, efficiency, volume and start-up time of 1kW class fuel processor showed 73 %(LHV), < 60 l and < 60 min, respectively. Advanced fuel processor showed more highly efficiency and shorter start-up time due to the improvement of heat exchanger and operating method. 1 kW and 3 kW class fuel processors have been evaluated for reliability and durability including with on/off test of developed catalysts and burner.

• 시스템엔지니어링학술지
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• 제18권2호
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• pp.94-107
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• 2022

Accidents prevention and mitigation is the highest priority of nuclear power plant (NPP) operation, particularly in the aftermath of the Fukushima Daiichi accident, which has reignited public anxieties and skepticism regarding nuclear energy usage. To deal with accident scenarios more effectively, operators must have ample and precise information about key safety parameters as well as their future trajectories. This work investigates the potential of machine learning in forecasting NPP response in real-time to provide an additional validation method and help reduce human error, especially in accident situations where operators are under a lot of stress. First, a base-case SGTR simulation is carried out by the best-estimate code RELAP5/MOD3.4 to confirm the validity of the model against results reported in the APR1400 Design Control Document (DCD). Then, uncertainty quantification is performed by coupling RELAP5/MOD3.4 and the statistical tool DAKOTA to generate a large enough dataset for the construction and training of neural-based machine learning (ML) models, namely LSTM, GRU, and hybrid CNN-LSTM. Finally, the accuracy and reliability of these models in forecasting system response are tested by their performance on fresh data. To facilitate and oversee the process of developing the ML models, a Systems Engineering (SE) methodology is used to ensure that the work is consistently in line with the originating mission statement and that the findings obtained at each subsequent phase are valid.

• 비파괴검사학회지
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• 제28권1호
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• pp.9-15
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• 2008

증기발생기 전열관에 생성되는 투자율변화에 의해 야기되는 신호왜곡 문제를 해결하기 위한 새로운 탐촉자를 개발하였다. 고리 1 호기 폐전열관에 생성된 자성상을 분리하여 자기이력곡선을 측정하였으며, 자성상이 생성되는 원인을 규명하기 위하여 고온 인장 시험을 수행하였다. 전산모사를 이용하여 탐촉자의 자정상 탐지조건을 결정하였으며, 디지털 신호전송을 위하여 신호처리용 전자회로를 소형화하여 탐촉자 속에 삽입하였다. 본 연구에서 개발된 신형 탐촉자를 이용하여 PVC 신호와 니켈 슬리빙 부위의 결함을 측정하였다. 신형 탐촉자는 보빈 탐촉자와 같이 고속으로 결함을 측정 할 수 있으며, 증기발생기 전열관의 탐상속도와 결함 탐지의 신뢰성을 증진시킬 수 있다.

• Nuclear Engineering and Technology
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• 제43권2호
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• pp.187-194
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• 2011

Korea Nuclear Fuel is developing the X-GEN fuel which shows high performance and robust reliability for the worldwide supply. However, the simplified code systems such as CESEC-III which were developed in 1970s are still used in the current Non-LOCA safety analysis of OPR1000 and APR1400 plants. Therefore, it is essential to secure an advanced safety analysis methodology to make the best use of the merits of X-GEN fuel. To accomplish this purpose, the $\b{i}$ntegrated $\b{s}$afety $\b{a}$nalysis $\b{m}$ethodology (iSAM), is developed by selecting the best-estimate thermal-hydraulic code RETRAN. iSAM possesses remarkable advantages, such as generality, integrity, and designer-friendly features. That is, iSAM can be applied to both OPR1000 and APR1400 plants and uses only one computer code, RETRAN, in the whole scope of the non-LOCA safety analyses. Also the iSAM adopts the unique and automatic initialization and run tool, $\b{a}$utomatic $\b{s}$teady-$\b{s}$tate $\b{i}$nitialization and $\b{s}$afety analysis too l (ASSIST), to enable unhandy designers to use the new design code RETRAN without difficulty. In this paper, a brief overview of the iSAM is given, and the results of applying the iSAM to typical non-LOCA transients being checked during the reload design are reported. The typical non-LOCA transients selected are the single control element assembly withdrawal (SCEAW) accident, the asymmetric steam generator transients (ASGT), the locked rotor (LR) accident, and bank CEA withdrawal (BCEAW) event. Comparison to current licensing results shows a close resemblance; thus, it reveals that the iSAM can be applied to the non-LOCA safety analysis of OPR1000 and APR1400 plants.

• 비파괴검사학회지
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• 제35권1호
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• pp.61-67
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• 2015

최근 국내 증기발생기 Alloy 600HTMA 전열관의 관 지지판 부위 외면 축균열 결함의 생성이 지속적으로 증가하고 있다. 이로 인하여 증기발생기가 설계수명 이전에 조기 교체되었으며 또는 교체 예정이다. 전열관 외면 축균열은 건전성 관리에 가장 위협이 되는 요소이므로 정밀한 건전성 평가가 요구된다. 와전류검사(ECT, eddy currunt testing)는 주기적으로 수행되어 지며 이 결과는 건전성 평가 입력 자료로 활용된다. ECT 검사시스템의 신뢰성은 검사기술과 평가자 기량에 의존하며, NDE 시스템 성능을 보여주는 지수는 열화탐지와 크기 측정 오차이다. 본 연구에서는 국내 평가자 성능이 반영된 크기 측정 오차와 그리고 최적의 균열 크기 측정 방법을 제시하였다. 실험은 국내 각기 다른 5개 회사에서 10명의 평가자가 참여한 다자간 비교시험의 결과를 사용하여 이루어졌다. 실험 결과 분석은 파괴검사 결과값과 비파괴검사로 측정된 값의 상관관계를 회귀분석을 통하여 이루어졌다.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.803-811
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• 2024

Nuclear accidents such as Fukushima Daiichi have highlighted the potential of passive safety systems to replace or complement active safety systems as part of the overall prevention and/or mitigation strategies. In addition, passive systems are key features of Small Modular Reactors (SMRs), for which they are becoming almost unavoidable and are part of the basic design of many reactors available in today's nuclear market. Nevertheless, their potential to significantly increase the safety of nuclear power plants still needs to be strengthened, in particular the ability of computer codes to determine their performance and reliability in industrial applications and support the safety demonstration. The PASTELS project (September 2020-February 2024), funded by the European Commission "Euratom H2020" programme, is devoted to the study of passive systems relying on natural circulation. The project focuses on two types, namely the SAfety COndenser (SACO) for the evacuation of the core residual power and the Containment Wall Condenser (CWC) for the reduction of heat and pressure in the containment vessel in case of accident. A specific design for each of these systems is being investigated in the project. Firstly, a straight vertical pool type of SACO has been implemented on the Framatome's PKL loop at Erlangen. It represents a tube bundle type heat exchanger that transfers heat from the secondary circuit to the water pool in which it is immersed by condensing the vapour generated in the steam generator. Secondly, the project relies on the CWC installed on the PASI test loop at LUT University in Finland. This facility reproduces the thermal-hydraulic behaviour of a Passive Containment Cooling System (PCCS) mainly composed of a CWC, a heat exchanger in the containment vessel connected to a water tank at atmospheric pressure outside the vessel which represents the ultimate heat sink. Several activities are carried out within the framework of the project. Different tests are conducted on these integral test facilities to produce new and relevant experimental data allowing to better characterize the physical behaviours and the performances of these systems for various thermo-hydraulic conditions. These test programmes are simulated by different codes acting at different scales, mainly system and CFD codes. New "system/CFD" coupling approaches are also considered to evaluate their potential to benefit both from the accuracy of CFD in regions where local 3D effects are dominant and system codes whose computational speed, robustness and general level of physical validation are particularly appreciated in industrial studies. In parallel, the project includes the study of single and two-phase natural circulation loops through a bibliographical study and the simulations of the PERSEO and HERO-2 experimental facilities. After a synthetic presentation of the project and its objectives, this article provides the reader with findings related to the physical analysis of the test results obtained on the PKL and PASI installations as well an overall evaluation of the capability of the different numerical tools to simulate passive systems.

• 비파괴검사학회지
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• 제24권4호
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• pp.384-389
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• 2004

원자력발전소 운전에 따른 경년열화 등에 의하여 원자력발전소 주요 기기 및 재료 등에 손상 발생 가능성이 있어 원자력법 및 관련 기술기준에서는 비파괴검사 방법을 이용하여 원자력발전소 주요 기기 및 배관의 용접부 등 취약 부위에 대한 건전성을 주기적으로 평가토록하고 있다. 이에 따라, 영광 6호기 가동중검사는 기기, 배관 및 구조물 비파괴검사, 압력용기 자동 초음파탐상검사, 원자로 내부 구조물 육안검사 및 증기발생기 전열관 와전류탐상검사로 구분하여 수행하였다. 원자력발전소 계통의 주요기기에 대한 비파괴검사 결과, 기기, 배관 및 구조물과 원자로 압력용기 용접부에 대해서는 특이 사항 발생 없이 적용 규격에 만족되고 건전한 것으로 최종 평가되었다. 특히, 배관 용접부에 대한 초음파탐상검사는 영광 5호기에서와 마찬가지로 ASME Code Sec. XI 1995년도 판에 따라 기량검증(Performance Demonstration : PD) 방법을 적용함으로써 검사 신뢰도를 확보하였다는데 큰 의미가 있다.

• Journal of Radiation Protection and Research
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• 제38권4호
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• pp.228-233
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• 2013

세계적으로 원전의 가동 년수 증가로 인하여 증기발생기와 같은 중요 설비의 교체가 지속적으로 이루어지고 있으며, 해체 시에는 대량의 방사성 금속 폐기물이 일시에 발생한다. 이러한 방사성 폐기물을 규제해제 후에 재활용하기 위해서는 정확한 잔류방사능을 측정하여야 한다. 그러나, 원자력시설에서 발생되는 금속 폐기물은 형상이 복잡하고, 재질별 특성이 다양하기 때문에 잔류방사능을 정확히 측정하기가 어렵다. 본 연구에서는 방사성 금속 폐기물의 정확한 잔류방사능을 측정하기 위한 절차를 수립하였고, 오염 대상 선원항 평가, 시료 대표성 확보 방안, 대면적 오염도 측정 장치 제작 및 밀도에 의한 자체흡수 보정인자 등을 평가하였다. 특히, 복잡한 구조의 금속 폐기물에 대하여 시료의 대표성을 확보하기 위하여 용융시킨 후 단순한 형태의 시료를 제조하였으며, 금속의 밀도 차이에 따른 보정인자를 결정하여 방사능 측정 결과의 신뢰성을 향상시켰다.