• Journal of the Computational Structural Engineering Institute of Korea
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• v.24 no.3
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• pp.343-351
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• 2011

Nuclear containment building is the last barrier for being secure from any nuclear power plant accident. Therefore, it is very important to understand the ultimate capacity of nuclear containment building to loads associated with severe accidents. LOCA (loss of coolant accident) is considered as the basic accidental load and CANDU-type containment building is considered as a target structure in order to conduct the numerical analysis for the structural safety of a containment building. The CANDU-type containment building is a prestressed concrete shell structure which has the dome and the cylindrical wall and is reinforced with bonded tendons. In this paper, the evaluation of ultimate internal pressure capacity was carried out by nonlinear analysis of a prestressed concrete containment building using 3-dimensional structural analysis system.

• Journal of the Earthquake Engineering Society of Korea
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• v.25 no.2
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• pp.53-58
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• 2021

Nuclear power plant's safety against seismic events is evaluated as risk values by probabilistic seismic safety assessment. The risk values vary by the seismic failure correlation between the structures, systems, and components (SSCs). However, most probabilistic seismic safety assessments idealized the seismic failure correlation between the SSCs as entirely dependent or independent. Such a consideration results in an inaccurate assessment result not reflecting real physical phenomenon. A nuclear power plant's seismic risk should be calculated with the appropriate seismic failure correlation coefficient between the SSCs for a reasonable outcome. An accident scenario that has an enormous impact on a nuclear power plant's seismic risk was selected. Moreover, the probabilistic seismic response analyses of a nuclear power plant were performed to derive appropriate seismic failure correlations between SSCs. Based on the analysis results, the seismic failure correlation coefficient between SSCs was derived, and the seismic fragility curve and core damage frequency of the loss of essential power event were calculated. Results were compared with the seismic fragility and core damage frequency of assuming the seismic failure correlations between SSCs were independent and entirely dependent.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4534-4550
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• 2022

Integrated severe accident codes should be capable of simulating not only specific physical phenomena but also entire plant behaviors, and in a sufficiently fast time. However, significant uncertainty may exist owing to the numerous parametric models and interactions among the various phenomena. The primary objectives of this study are to present best-practice uncertainty and sensitivity analysis results regarding the evolutions of severe accidents (SAs) and fission product source terms and to determine the effects of mitigation measures on them, as expected during a short-term station blackout (STSBO) of a reference pressurized water reactor (optimized power reactor (OPR)1000). Three reference scenarios related to the STSBO accident are considered: one base and two mitigation scenarios, and the impacts of dedicated severe accident mitigation (SAM) actions on the results of interest are analyzed (such as flammable gas generation). The uncertainties are quantified based on a random set of Monte Carlo samples per case scenario. The relative importance values of the uncertain input parameters to the results of interest are quantitatively evaluated through a relevant sensitivity/importance analysis.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.673-682
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• 2018

One of the most widely used methods to estimate core damage during a nuclear power plant accident is containment radiation measurement. The evolution of severe accidents is extremely complex, leading to uncertainty in the containment dose rate (CDR). Therefore, it is difficult to accurately determine core damage. This study proposes to conduct uncertainty analysis of CDR for core damage assessment. First, based on source term estimation, the Monte Carlo (MC) and point-kernel integration methods were used to estimate the probability density function of the CDR under different extents of core damage in accident scenarios with late containment failure. Second, the results were verified by comparing the results of both methods. The point-kernel integration method results were more dispersed than the MC results, and the MC method was used for both quantitative and qualitative analyses. Quantitative analysis indicated a linear relationship, rather than the expected proportional relationship, between the CDR and core damage fraction. The CDR distribution obeyed a logarithmic normal distribution in accidents with a small break in containment, but not in accidents with a large break in containment. A possible application of our analysis is a real-time core damage estimation program based on the CDR.

• Nuclear Engineering and Technology
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• v.27 no.1
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• pp.33-44
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• 1995

Public acceptance has become an important factor in nuclear power program particularly after Chernobyl accident and recent rapid democratization in Korea. Methods reflecting public opinions in order to improve public acceptance are firstly to understand what the public think about nuclear power plant and secondly to find out the public preference values for its policies. For this purpose, simplified multi-attribute utility (MAU) model was applied to analyze the public perception pattern for fire power production systems. And the conjoint analysis was applied to find out the quantitative values of public preferences for twelve policy alternatives to improve the safety and to support communities surrounding nuclear power plants in Korea. To implement these perception and preference analyses, mail survey was conducted to the Qualified sample who had the experience of visiting nuclear power plant. Diagnosis of their perception pattern for five power production systems was made by the simplified MAU model. Estimation of the quantitative preference values for potential policy alternatives was made by the conjoint measurement technique, which made it possible to forecast the effectiveness of each option. The results from the qualified sample and the methods used in this study would be helpful to set up new policy of nuclear power plant.

• Journal of the Korean Society of Safety
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• v.26 no.5
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• pp.105-109
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• 2011

Since a severe accident, which happens in low frequency, can cause serious damages, the interests in off-site consequence analysis for a nuclear power plant have been increased after Chernobyl, TMI and Fukushima accidents. Consequences, which are the effects on health and environment caused by released radioisotopes, are evaluated using MACCS II code based on the method of Level 3 PSA. To perform a consequence analysis for the reference plants, the input data of the code were generated such as meteorological data, population distribution, release fractions, and so on. Using these input data, acute and lifetime dose as an organ, CCDF for early fatalities and latent cancer fatalities, and average individual risk were analyzed by using MACCS II code in this study. These results might contribute to establishing accident management plan and quantitative health object.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2821-2835
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• 2020

The phenomenon identification and ranking table (PIRT) is an important basis in the nuclear power plant (NPP) thermal-hydraulic analysis. This study focuses on the importance ranking of the input parameters when lacking the PIRT, and the target scenario is the small break loss of coolant accident (SBLOCA) in a pressurized water reactor (PWR) CPR1000. A total of 54 input parameters which might have influence on the figure of merit (FOM) were identified, and the sensitivity measure of each input on the FOM was calculated through an optimized moment-independent global sensitivity analysis method. The importance ranking orders of the parameters were transformed into the Savage scores, and the parameters were categorized based on the Savage scores. A parameter importance ranking table for the SBLOCA scenario of the CPR1000 reactor was obtained, and the influences of some important parameters at different break sizes and different accident stages were analyzed.

• Nuclear Engineering and Technology
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• v.41 no.7
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• pp.945-952
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• 2009

The Three Mile Island Unit 2 (TMI-2) accident provides unique full scale data, thus providing opportunities to check the capability of codes to model overall plant behavior and to perform a spectrum of sensitivity and uncertainty calculations. As part of the TMI-2 analysis benchmark exercise sponsored by the Organization for Economic Cooperation and Development Nuclear Energy Agency (OECD NEA), several member countries are continuing to improve their system analysis codes using the TMI-2 data. The Republic of Korea joined this benchmark exercise in November 2005. Seoul National University has analyzed the TMI-2 accident as well as the currently proposed alternative scenario along with a sensitivity study using the Modular Accident Analysis Program Version 4.03 (MAAP4.03) code in collaboration with the Korea Hydro and Nuclear Power Company. Two input files are required to simulate the TMI-2 accident with MAAP4: the parameter file and an input deck. The user inputs various parameters, such as volumes or masses, for each component. The parameter file contains the information on TMI-2 relevant to the plant geometry, system performance, controls, and initial conditions used to perform these benchmark calculations. The input deck defines the operator actions and boundary conditions during the course of the accident. The TMI-2 accident analysis provided good estimates of the accident output data compared with the OECD TMI-2 standard reference. The alternative scenario has proposed the initial event as a loss of main feed water and a small break on the hot leg. Analysis is in progress along with a sensitivity study concerning the break size and elevation.

• Journal of the Korean Society of Safety
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• v.26 no.5
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• pp.99-104
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• 2011

An ILRT Interval for a nuclear power plant in Korea was extended from once in five years to once in ten years. Therefore, it is necessary to evaluate risk impact for ILRT interval extensions. In this paper, input data were generated for the reference plants, KSNP, using raw data such as meteorological data, population distribution data and source term data. And, using MACCS II code the risk impact assessment was performed based on the two methodologies of NUREG-1493 and NEI Interim Report. The risk impact derived from an ILRT interval extension was identified not to be significant. It is considered to apply this study and results to making an accident management plan and safety goal, and to the field of public acceptance.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.366-372
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• 2020

Using several accessible published data sets, we analyzed the temporal change of ¹³⁷Cs radioactivity (per unit mass of sample) in marine sediments and investigated the effect of the water content of sediment on the ¹³⁷Cs radioactivity, to understand the behavior of ¹³⁷Cs present in marine environments. The ¹³⁷Cs radioactivity in sediments decreased more slowly in the Baltic Sea (near the Chernobyl nuclear power plant) than in the ocean near the Fukushima Daiichi nuclear power plant (FDNPP). The ¹³⁷Cs radioactivity in the sediment near the FDNPP tended to increase as the water content increased, and the water content decreased at certain sampling sites near the FDNPP for several years. Additionally, the decrease in the water content contributed to 51.2% of the average ¹³⁷Cs radioactivity decrease rate for the same period. Thus, it may be necessary to monitor both the ¹³⁷Cs radioactivity and the water content for marine sediments to track the ¹³⁷Cs that was discharged from the sites of Chernobyl and Fukushima nuclear power plants where severe accidents occurred.