• Journal of the Korean Society of Safety
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• v.27 no.5
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• pp.219-223
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• 2012

For an off-site consequence analysis at nuclear power plant, MELCOR Accident Consequence Code System(MACCS) II code is widely used as a software tool. In this study, the algorithm of web-based off-site consequence analysis program(OSCAP) using the MACCS II code was developed for an Integrated Leak Rate Test (ILRT) interval extension and Level 3 probabilistic safety assessment(PSA), and verification and validation(V&V) of the program was performed. The main input data for the MACCS II code are meteorological, population distribution and source term information. However, it requires lots of time and efforts to generate the main input data for an off-site consequence analysis using the MACCS II code. For example, the meteorological data are collected from each nuclear power site in real time, but the formats of the raw data collected are different from each site. To reduce the efforts and time for risk assessments, the web-based OSCAP has an automatic processing module which converts the format of the raw data collected from each site to the input data format of the MACCS II code. The program also provides an automatic function of converting the latest population data from Statistics Korea, the National Statistical Office, to the population distribution input data format of the MACCS II code. For the source term data, the program includes the release fraction of each source term category resulting from modular accident analysis program(MAAP) code analysis and the core inventory data from ORIGEN. These analysis results of each plant in Korea are stored in a database module of the web-based OSCAP, so the user can select the defaulted source term data of each plant without handling source term input data.

• Journal of Ocean Engineering and Technology
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• v.30 no.6
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• pp.451-457
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• 2016

For a carferry with a displacement of 1,633 tonf, a seakeeping analysis-based direct load approach (DLA) was used in Part I of these series, where the final deliverable was the long-term probabilistic acceleration components. In Part II of these series, the tangential acceleration components are explained based on two approaches: a standard called the IMO CSS code and simple formulas with the probable maximum roll and pitch rotations. The subsequent tangential acceleration-induced external force components are also introduced for these two approaches. The lashing strength components were selected from the IMO CSS code. It was assumed that two different vehicles (a car and a truck) were stowed at the most distant locations on the main deck to assume the largest tangential acceleration components and were secured with four steel wires with longitudinal and transverse lashing angles of $45^{\circ}$. Four cases were considered, with different methods for predicting the acceleration components and different tools for the external loads and lashing strengths involved: cases Rule-LS (rule-based maximum probable roll and pitch angles for predicting the acceleration components in conjunction with LashingSafety), DLA-LS (seakeeping-based long-term acceleration components with LashingSafety), CSS-LC (IMO CSS code-based acceleration components using LashCon), and CSS-LS (IMO CSS code-based acceleration components using LashingSafety). In terms of the acceleration and external force components, the CSS-LC and CSS-LS results are more than two times the results of Rule-LS. Thus, when the external forces and lashing strengths are evaluated using CSS-LC and CSS-LS, the truck needs more lashing wires, while Rule-LS and DLA-LS predict that the present lashing configuration is on the safe side.

• Tunnel and Underground Space
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• v.14 no.3
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• pp.203-214
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• 2004

For total system performance of a potential radwaste repository, a hypothetical site is assumed with feasible boundary conditions. Assuming a coastal repository, the sensitivity of the depth and the location of a repository along with the distance to a joint on groundwater transport pathways is studied. Results from Connectflow analysis could be used as input of the MASCOT-t the probabilistic safety assessment code.

• Structural Engineering and Mechanics
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• v.85 no.6
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• pp.793-808
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• 2023

Türkiye is located in a region where destructive earthquakes are frequently experienced due to its geological characteristics and geographical location. Therefore, considering the possibility of a devastating earthquake at any time, determining the reinforced concrete (RC) building seismic safety, constructed before or after the current seismic buildings code, is one of the most important issues to be completed firstly. For this purpose, rapid assessment methods developed to quickly determine the seismic safety of buildings are available in the literature. Comparison of the principles of Principles of the Determination of Risky Structures-2019, Column and Wall Index Method, P25 Scoring Method and Improved Discriminant Analysis Method, which are among these methods, have been aimed within the scope of this study. Within the scope of this paper, a total of 43 buildings in the Yalova/Çınarcık region of Türkiye that the damage level was determined by street observation method immediately after the 1999 Kocaeli (Izmit) Earthquake; 15 buildings with heavy damage and 28 buildings with moderate damage were examined by rapid assessment methods. Although the risk detection difference was not separated as a clear line in any of the methods used, the results obtained from the rapid assessment methods are evaluated as being compatible with the detected after earthquake structural seismic behavior of the buildings. The PDRS-2019 and column and wall index method gave the most approximate results. In the results obtained from the analyzes; structural features such as number of floors, frame continuity, soft/weak story irregularity, effective shear strength area, existence of heavy overhangs in plan, type of structural system have been found to be significantly effective on the earthquake behavior of buildings.

• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.684-701
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• 2016

The Gaussian process model (GPM) is a flexible surrogate model that can be used for nonparametric regression for multivariate problems. A unique feature of the GPM is that a prediction variance is automatically provided with the regression function. In this paper, we estimate the safety margin of a nuclear power plant by performing regression on the output of best-estimate simulations of a large-break loss-of-coolant accident with sampling of safety system configuration, sequence timing, technical specifications, and thermal hydraulic parameter uncertainties. The key aspect of our approach is that the GPM regression is only performed on the dominant input variables, the safety injection flow rate and the delay time for AC powered pumps to start representing sequence timing uncertainty, providing a predictive model for the peak clad temperature during a reflood phase. Other uncertainties are interpreted as contributors to the measurement noise of the code output and are implicitly treated in the GPM in the noise variance term, providing local uncertainty bounds for the peak clad temperature. We discuss the applicability of the foregoing method to reduce the use of conservative assumptions in best estimate plus uncertainty (BEPU) and Level 1 probabilistic safety assessment (PSA) success criteria definitions while dealing with a large number of uncertainties.

• Nuclear Engineering and Technology
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• v.28 no.6
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• pp.507-521
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• 1996

Quantification of uncertainties in the source term estimations by a large computer code, such as MELCOR and MAAP, is an essential process of the current Probabilistic safety assessment. The main objective of the present study is to investigate the applicability of a combined procedure of the response surface method (RSM) based on input determined from a statistical design and the Latin hypercube sampling (LHS) technique for the uncertainty analysis of CsI release fractions under a Hypothetical severe accident sequence of a station blackout at Younggwang nuclear power plant using MAAP3. OB code as a benchmark problem. On the basis of the results obtained in the present work, the RSM is recommended to be used as a principal tool for an overall uncertainty analysis in source term quantifications, while using the LHS in the calculations of standardized regression coefficients (SRC) and standardized rank regression coefficient (SRRC) to determine the subset of the most important input parameters in the final screening step and to check the cumulative distribution functions obtained by RSM. Verification of the response surface model for its sufficient accuracy is a prerequisite for the reliability of the final results that can be obtained by the combined procedure proposed in the present work.

• Journal of the Korean Society of Safety
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• v.28 no.3
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• pp.123-128
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• 2013

The Advanced Power Reactor Plus(APR+), which is a GEN III+ reactor based on the APR1400, is being developed in Korea. In order to enhance the safety of the APR+, a passive auxiliary feedwater system(PAFS) has been adopted in the APR+. The PAFS replaces the conventional active auxiliary feedwater system(AFWS) by introducing a natural driving force mechanism while maintaining the system function of cooling the primary side and removing the decay heat. As the PAFS completely replaces the conventional AFWS, it is required to verify the cooling capacity of PAFS for the core damage frequency(CDF) evaluation. For this reason, this paper discusses the cooling performance of the PAFS during transient accidents. The test case and scenarios were picked from the result of the sensitivity analysis in APR+ Probabilistic Safety Assessment(PSA). The analysis was performed by the best estimate thermal-hydraulic code, RELAP5/.MOD3.3. This study shows that the plant maintains the stable state without the core damages under the given test scenarios. The results of PSA considering this analysis' results shows that the CDF values are decreased. The analysis results can be used for more realistic and accurate performance of a PSA.

• Earthquakes and Structures
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• v.19 no.6
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• pp.471-484
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• 2020

Previous earthquakes show that, structural safety evaluations should include the evaluation of nonstructural components. Failure of nonstructural components can affect the operational capacity of critical facilities, such as hospitals and fire stations, which can cause an increase in number of deaths. Additionally, failure of nonstructural components may result in economic, architectural, and historical losses of community. Accelerations and random vibrations must be under the predefined limitations in structures with high technological equipment, data centers in this case. Failure of server equipment and anchored server racks are investigated in this study. A probabilistic study is completed for a low-rise rigid sample structure. The structure is investigated in two versions, (i) conventional fixed-based structure and (ii) with a base isolation system. Seismic hazard assessment is completed for the selected site. Monte Carlo simulations are generated with selected parameters. Uncertainties in both structural parameters and mechanical properties of isolation system are included in simulations. Anchorage failure and vibration failures are investigated. Different methods to generate fragility curves are used. The site-specific annual hazard curve is used to generate risk curves for two different structures. A risk matrix is proposed for the design of data centers. Results show that base isolation systems reduce the failure probability significantly in higher floors. It was also understood that, base isolation systems are highly sensitive to earthquake characteristics rather than variability in structural and mechanical properties, in terms of accelerations. Another outcome is that code-provided anchorage failure limitations are more vulnerable than the random vibration failure limitations of server equipment.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.759-764
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• 1997

본 논문에서는 영광 3,4호기의 초기 부분충수 운전중 정지냉각 상실 사건에 대하여 확률론적 안전성평가(Probabilistic Safety Assessment; PSA)를 수행하였다. 1단계 PSA 결과인 노심손상빈도에 크게 영향을 끼치는 인간행위는 THERP(technique for human error rate prediction)를 사용하여 평가하였고, 사고경위는 KIRAP(KAERI integrated reliability analysis code package)을 이용하여 정량화하였다. 영광 3,4호기의 부분충수 운전중 정지냉각 상실 사건에 대한 예비적인 PSA 결과, 노심손상 빈도는 1.43E-6로 평가되었고 노심손상 빈도에 주요하게 기여하는 것은 원자로 냉각재 보충에 대한 운전원의 진단 실패로 나타났다. 노심손상빈도를 감소하는 방안의 하나는 운전원의 진단오류 확률을 낮추기 위해 노심손상까지의 운전원 여유시간을 확장하는 것이다. 그러나 보다 구체적인 결과는 분석에 필요한 여러 가지 자료들을 검토하고 PSA를 다시 수행해야 얻을 수 있을 것으로 판단된다.

• Steel and Composite Structures
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• v.19 no.4
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• pp.811-827
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• 2015

This paper presents a probabilistic investigation of American and European specifications (i.e., AISC and Eurocode 4) for square concrete-filled steel tubular (CFT) stub columns. The study is based on experimental results of 100 axially loaded square CFT stub columns from the literature. By comparing experimental results for ultimate loads with code-predicted column resistances, the uncertainty of resistance models is analyzed and it is found that the modeling uncertainty parameter can be described using random variables of lognormal distribution. Reliability analyses were then performed with/without considering the modeling uncertainty parameter and the safety level of the specifications is evaluated in terms of sufficient and uniform reliability criteria. Results show that: (1) The AISC design code provided slightly conservative results of square CFT stub columns with reliability indices larger than 3.25 and the uniformness of reliability indices is no better because of the quality of the resistance model; (2) The uniformness of reliability indices for the Eurocode 4 was better than that of AISC, but the reliability indices of columns designed following the Eurocode 4 were found to be quite below the target reliability level of Eurocode 4.