• Nuclear Engineering and Technology
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• 제41권4호
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• pp.391-402
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• 2009

This paper introduces the development of safety assessment technology in Korea, focusing on the activities of the Korea Atomic Energy Research Institute in the areas of system thermal hydraulics, severe accidents and probabilistic safety assessment. In the 1970s and 1980s, safety analysis codes and methodologies were introduced from the United States, France, Canada and other developed countries along with technology related to the construction and operation of nuclear power plants. The main focus was on understanding and utilizing computer codes that were sourced from abroad up to the early 1990s, when efforts to develop domestic safety analysis codes and methodologies became active. Remarkable achievements have been made over the last 15 years in the development and application of safety analysis technologies. In addition, significant experimental work has been performed to verify the safety characteristics of reactors and fuels as well as to support the development and validation of analysis methods.

• Nuclear Engineering and Technology
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• 제48권6호
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• pp.1423-1432
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• 2016

The integrity of a reactor pressure vessel (RPV) related to pressurized thermal shocks (PTSs) has been extensively studied. This paper introduces an integrity assessment of an RPV subjected to a PTS transient based on the French codes. In the USA, the "screening criterion" for maximum allowable embrittlement of RPV material is developed based on the probabilistic fracture mechanics. However, in the French RCC-M and RSE-M codes, which are developed based on the deterministic fracture mechanics, there is no "screening criterion". In this paper, the methodology in the RCC-M and RSE-M codes, which are used for PTS analysis, are firstly discussed. The bases of the French codes are compared with ASME and FAVOR codes. A case study is also presented. The results show that the method in the RCC-M code that accounts for the influence of cladding on the stress intensity factor (SIF) may be nonconservative. The SIF almost doubles if the weld residual stress is considered. The approaches included in the codes differ in many aspects, which may result in significant differences in the assessment results. Therefore, homogenization of the codes in the long time operation of nuclear power plants is needed.

• 한국환경보건학회지
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• 제26권3호
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• pp.55-60
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• 2000

Wire codes, introduced by Wertheimer and Leeper, are interesting largely because several case-control studies have reported associations between them and certain types of cancers, including childhood leukemia. This paper is concerned with the reliability of wire codes as a surrogate for residential electric and magnetic field, divided into four categories based on the electrical wiring within 40m of homes. To examine the correlation between spot measurements and wire codes, EMF was measured by using HI-3604 and classified by wire codes. The correlation between long-transformed RMS of magnetic field and wire codes was 0.81. But the correlation between log-transformed RMS of electric field and wire codes was 0.27. The results led us to determine a four category from of the Wertheimer and Leeper code that 1) may be useful in domestic epidemiological studies and 2) is not suitable for electric field exposure. Thus, the critical point of exposure assessment in residential magnetic field is to classify a wire code precisely.

• 한국건축시공학회:학술대회논문집
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• 한국건축시공학회 2012년도 춘계 학술논문 발표대회
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• pp.151-152
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• 2012

Recently, there are growing interests in building LCCO₂ Assessment to reduce carbon emissions. However, existing methods of assessment system include inefficiency in the process of CO₂ calculation requiring considerable data input. Therefore, the purpose of this study is to develop an efficient building assessment system appropriate to material production in construction stage. To that end, quantity input technology was limited to data mapping. Also quantity calculation based on work breakdown structure and item codes consisted of hierarchical structure that is based on facet classification were analyzed. As a result, connectivity links of quantity calculation and CO₂ functional units through item codes for data mapping, and assessment system including calculation and database parts were developed.

• Computers and Concrete
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• 제2권1호
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• pp.31-53
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• 2005

In the present structural codes the safety verification is based on a linear analysis of the structure and the satisfaction of ultimate and serviceability limit states, using a semi-probabilistic security format through the consideration of partial safety factors, which affect the action values and the characteristic values of the material properties. In this context, if a non-linear structural analysis is wanted a difficulty arises, because the global safety coefficient, which could be obtained in a straightforward way from the non-linear analysis, is not directly relatable to the different safety coefficient values usually used for the different materials, as is the case for reinforced concrete structures. The work here presented aims to overcome this difficulty by proposing a methodology that generalises the format of safety verification based on partial safety factors, well established in structural codes within the scope of linear analysis, for cases where non-linear analysis is needed. The methodology preserves the principal assumptions made in the codes as well as a reasonable simplicity in its use, including a realistic definition of the material properties and the structural behaviour, and it is based on the evaluation of a global safety coefficient. Some examples are presented aiming to clarify and synthesise all the options that were taken in the application of the proposed methodology, namely how to transpose the force distributions obtained with a non-linear analysis into design force distributions. One of the most important features of the proposed methodology, the ability for comparing the simplified procedures for second order effects evaluation prescribed in the structural codes, is also presented in a simple and systematic way. The potential of the methodology for the development and assessment of alternative and more accurate procedures to those already established in codes of practice, where non-linear effects must be considered, is also indicated.

• 산업기술연구
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• 제29권B호
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• pp.173-180
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• 2009

Simulations using two well-known commercial codes, WAsP and WindSIM, were performed to predict the wind resources in complex terrains of Korea. The predictions from the codes were compared with the measured data. Cross predictions were performed for two closely located measurement sites. The results from WindSIM were found to be more accurate than those from WAsP. The predictions for wind velocity and direction in five different sites of complex terrain from WAsP and WindSIM were also compared. It was found that if the self prediction of the wind velocity and direction from WAsP is close to the measured wind data, the discrepancies between WAsP results and WindSIM results are also close.

• Communications for Statistical Applications and Methods
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• 제27권5호
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• pp.569-578
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• 2020

Various graph clustering methods have been introduced to identify communities in social or biological networks. This paper studies the entropy-based and the Markov chain-based methods in clustering the undirected graph. We examine the performance of two clustering methods with conventional methods based on quality measures of clustering. For the real applications, we collect the mathematical subject classification (MSC) codes of research papers from published mathematical databases and construct the weighted code-to-document matrix for applying graph clustering methods. We pursue to group MSC codes into the same cluster if the corresponding MSC codes appear in many papers simultaneously. We compare the MSC clustering results based on the several assessment measures and conclude that the Markov chain-based method is suitable for clustering the MSC codes.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3367-3378
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• 2021

The design phase and safety assessment of Generation IV liquid metal-cooled fast reactors calls for the improvement of fuel pin performance codes, in particular the enhancement of their predictive capabilities towards uranium-plutonium mixed oxide fuels and stainless-steel cladding under irradiation in fast reactor environments. To this end, the current capabilities of fuel performance codes must be critically assessed against experimental data from available irradiation experiments. This work is devoted to the assessment of three European fuel performance codes, namely GERMINAL, MACROS and TRANSURANUS, against the irradiation of two fuel pins selected from the SUPERFACT-1 experimental campaign. The pins are characterized by a low enrichment (~ 2 wt.%) of minor actinides (neptunium and americium) in the fuel, and by plutonium content and cladding material in line with design choices envisaged for liquid metal-cooled Generation IV reactor fuels. The predictions of the codes are compared to several experimental measurements, allowing the identification of the current code capabilities in predicting fuel restructuring, cladding deformation, redistribution of actinides and volatile fission products. The integral assessment against experimental data is complemented by a code-to-code benchmark focused on the evolution of quantities of engineering interest over time. The benchmark analysis points out the differences in the code predictions of fuel central temperature, fuel-cladding gap width, cladding outer radius, pin internal pressure and fission gas release and suggests potential modelling development paths towards an improved description of the fuel pin behaviour in fast reactor irradiation conditions.

• Structural Engineering and Mechanics
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• 제80권6호
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• pp.677-688
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• 2021

The building design codes and standards in many countries usually are either fully or partially adopted from the international codes. However, regional conditions like the quality of construction industry and different statistical parameters of load and resistance have essential roles in the code calibration of building design codes. This paper presents a probabilistic approach to assess the reliability level of adopted national building codes by simulating design situations and considering all load combinations. The impact of the uncertainty of wind and earthquake loads, which are entirely regional condition dependent and have a high degree of uncertainty, are quantified. In this study, the design situation is modeled by generating thousands of numbers for load effect ratios, and the reliability level of steel elements for all load combinations and different load ratios is established and compared to the target reliability. This approach is applied to the Iranian structural steel code as a case study. The results indicate that the Iranian structural steel code lacks safety in some load combinations, such as gravity and earthquake load combinations, and is conservative for other load combinations. The present procedure can be applied to the assessment of the reliability level of other national codes.

• Nuclear Engineering and Technology
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• 제55권3호
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• pp.884-894
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• 2023

Design and safety assessment of fuel pins for application in innovative Generation IV fast reactors calls for a dedicated nuclear fuel modelling and for the extension of the fuel performance code capabilities to the envisaged materials and irradiation conditions. In the INSPYRE Project, comprehensive and physics-based models for the thermal-mechanical properties of U-Pu mixed-oxide (MOX) fuels and for fission gas behaviour were developed and implemented in the European fuel performance codes GERMINAL, MACROS and TRANSURANUS. As a follow-up to the assessment of the reference code versions ("pre-INSPYRE", NET 53 (2021) 3367-3378), this work presents the integral validation and benchmark of the code versions extended in INSPYRE ("post-INSPYRE") against two pins from the SUPERFACT-1 fast reactor irradiation experiment. The post-INSPYRE simulation results are compared to the available integral and local data from post-irradiation examinations, and benchmarked on the evolution during irradiation of quantities of engineering interest (e.g., fuel central temperature, fission gas release). The comparison with the pre-INSPYRE results is reported to evaluate the impact of the novel models on the predicted pin performance. The outcome represents a step forward towards the description of fuel behaviour in fast reactor irradiation conditions, and allows the identification of the main remaining gaps.