• Nuclear Engineering and Technology
• /
• 제49권2호
• /
• pp.286-294
• /
• 2017

The events at the Fukushima Daiichi Nuclear Power Station in March 2011 point out, among other matters, that concurrent accidents at multiple units of a site can occur in reality. Although site risk has been deterministically considered to some extent in nuclear power plant siting and design, potential occurrence of multi-unit accident sequences at a site was not investigated in sufficient detail thus far in the nuclear power community. Therefore, there is considerable worldwide interest and research effort directed toward multi-unit site risk assessment, especially in the countries with high-density nuclear-power-plant sites such as Korea. As the technique of probabilistic safety assessment (PSA) has been successfully applied to evaluate the risk associated with operation of nuclear power plants in the past several decades, the PSA having primarily focused on single-unit risks is now being extended to the multi-unit PSA. In this paper we first characterize the site risk with explicit consideration of the risk associated with spent fuel pools as well as the reactor risks. The status of multi-unit risk assessment is discussed next, followed by a description of the emerging issues relevant to the multi-unit risk evaluation from a practical standpoint.

• 한국안전학회지
• /
• 제35권1호
• /
• pp.97-106
• /
• 2020

The objective of this paper is to suggest a new quantification method for multi-unit probabilistic safety assessment (PSA) that removes the overestimation error caused by the existing delete-term approximation (DTA) based quantification method. So far, for the actual plant PSA model quantification, a fault tree with negates have been solved by the DTA method. It is well known that the DTA method induces overestimated core damage frequency (CDF) of nuclear power plant (NPP). If a PSA fault tree has negates and non-rare events, the overestimation in CDF drastically increases. Since multi-unit seismic PSA model has plant level negates and many non-rare events in the fault tree, it should be very carefully quantified in order to avoid CDF overestimation. Multi-unit PSA fault tree has normal gates and negates that represent each NPP status. The NPP status means core damage or non-core damage state of individual NPPs. The non-core damage state of a NPP is modeled in the fault tree by using a negate (a NOT gate). Authors reviewed and compared (1) quantification methods that generate exact or approximate Boolean solutions from a fault tree, (2) DTA method generating approximate Boolean solution by solving negates in a fault tree, and (3) probability calculation methods from the Boolean solutions generated by exact quantification methods or DTA method. Based on the review and comparison, a new intersection removal by probability (IRBP) method is suggested in this study for the multi-unit PSA. If the IRBP method is adopted, multi-unit PSA fault tree can be quantified without the overestimation error that is caused by the direct application of DTA method. That is, the extremely overestimated CDF can be avoided and accurate CDF can be calculated by using the IRBP method. The accuracy of the IRBP method was validated by simple multi-unit PSA models. The necessity of the IRBP method was demonstrated by the actual plant multi-unit seismic PSA models.

• 한국주거학회논문집
• /
• 제19권2호
• /
• pp.43-54
• /
• 2008

The aim of the research is to find out the direction of unit-plans in multi-dwellings for the future society. Shinonome Canal Court where residents actually live now are the objects in this study, and the residential floor plan of unit-plans were analyzed to find out the typical types. The analysis was focused on the unit-plans of 5 blocks of Shinonome Canal Court. Space Syntax Theory was used as the analysis method. As the first stage of the analysis, justified graphs were made to find out the characters of unit-plans through the classification of the graphs. Contents of the analysis are as follows: Relationship between classified justified graphs and dimension according to node number. Relationship between classified justified graph patterns and unit-plans. Characters of unit-plans in each blocks. Shinonome Canal Court consists of mainly small scale unit-plans and 30unit-plans are classified. 1LDK, 2LDK, 1LDK+S, 1LDK+f are typical unit-plans which are mainly supplied in the complex. It is noted that the results of the analysis by node, justified graph pattern and dimension are the same. It also presents diverse unit-plans which shows a change in nLDK pattern or add f (foyer), AN (annex), S (service room), Fs (free space) to basic nLDK type. In summary, it demonstrates the possibility of creating new residental floor plans in multi-dwellings.

• Nuclear Engineering and Technology
• /
• 제50권8호
• /
• pp.1210-1216
• /
• 2018

Site or multi-unit (MU) risk assessment has been a major issue in the field of nuclear safety study since the Fukushima accident in 2011. There have been few methods or experiences for MU risk assessment because the Fukushima accident was the first real MU accident and before the accident, there was little expectation of the possibility that an MU accident will occur. In addition to the lack of experience of MU risk assessment, since an MU nuclear power plant site is usually very complex to analyze as a whole, it was considered that a systematic method such as probabilistic safety assessment (PSA) is difficult to apply to MU risk assessment. This paper proposes a new MU risk assessment methodology by using the conventional PSA methodology which is widely used in nuclear power plant risk assessment. The logical failure structure of a site with multiple units is suggested from the definition of site risk, and a decomposition method is applied to identify specific MU failure scenarios.

• Nuclear Engineering and Technology
• /
• 제53권3호
• /
• pp.967-973
• /
• 2021

In 2011, an earthquake and subsequent tsunami hit the Fukushima Daiichi Nuclear Power Plant, causing simultaneous accidents in several reactors. This accident shows us that if there are several reactors on site, the seismic risk to multiple units is important to consider, in addition to that to single units in isolation. When a seismic event occurs, a seismic-failure correlation exists between the nuclear power plant's structures, systems, and components (SSCs) due to their seismic-response and seismic-capacity correlations. Therefore, it is necessary to evaluate the multi-unit seismic risk by considering the SSCs' seismic-failure-correlation effect. In this study, a methodology is proposed to obtain the seismic-response-correlation coefficient between SSCs to calculate the risk to multi-unit facilities. This coefficient is calculated from a probabilistic multi-unit seismic-response analysis. The seismic-response and seismic-failure-correlation coefficients of the emergency diesel generators installed within the units are successfully derived via the proposed method. In addition, the distribution of the seismic-response-correlation coefficient was observed as a function of the distance between SSCs of various dynamic characteristics. It is demonstrated that the proposed methodology can reasonably derive the seismic-response-correlation coefficient between SSCs, which is the input data for multi-unit seismic probabilistic safety assessment.

• 품질경영학회지
• /
• 제17권1호
• /
• pp.1-10
• /
• 1989

In general, the characteristics of components which consist of multi-component system can not be the same. This paper proposes a maintenance model of multi-component system considering the characteristics of each component. In this paper, multi-component system is divided into three components-critical unit, major unit and minor unit, respectively. This paper determines the optimal replacement time of the system which minimizes total maintenance cost, optimal replacement period of major unit and initial stock quantity of minor unit within this optimal replacement time. Numerical examples are shown when the failure times of each unit have gamma distribution.

• 한국정밀공학회:학술대회논문집
• /
• 한국정밀공학회 2003년도 춘계학술대회 논문집
• /
• pp.1046-1050
• /
• 2003

In this paper, we describe study on the development or multi-path ultrasonic gas flowmeter using a transit time method. This system includes 5 pairs of ultrasonic transducers. ultrasonic signal processing unit using switch matrix method, computation algorithm of gas flow rate, spool piece type multi-path pipe unit. We have developed enhanced type of main ultrasonic signal processing unit using switch matrix method fer multi-path ultrasonic gas flowmeter. Also, we have developed the new transmitting ＆ receiving method of ultrasonic waves and the new signal processing algorithm for the computation of ultrasonic transit time from received ultrasonic waves. In this study, we have designed more compact signal processing unit compared with the conventional hardware system of multi-path ultrasonic gas flowmeter. We have confirmed its reliability for multi-path ultrasonic gas flowmeter through the laboratory test using calibration system. In the future. we will perform the field test for the developed system in the POSCO gas line.

• 한국기계가공학회지
• /
• 제11권6호
• /
• pp.9-15
• /
• 2012

Vibrational film has been more employed in ear-phones or small type of speakers along with a wide use of portable multi-media equipments such as MP3 and MP4. However, the current hand work production process of diaphragms is inefficient. In this study, a die-and-mold and a single production forming machine are developed, and they result in a multi-production forming machine. The multi-production forming machine consists primarily of a film feeding unit and an unwinding unit. A vacuum suction device provides the film feeding unit, while the unwinding unit is obtained using an appropriate damper. The advantage of the developed single production forming machine is shown according to a proper voice test.

• 한국기계가공학회지
• /
• 제13권1호
• /
• pp.52-58
• /
• 2014

This study consists of two parts. The first discusses the development of a single production forming machine which was reported in earlier papers. The second outlines the development of a multi-production forming machine, which consists primarily of a film feeding unit, an unwinding unit, and a heating block unit. The heating block unit of the multi-production forming machine has 30 members per die. An analysis of the stress deformation and temperature deviation of this machine is carried out using ANSYS Workbench and CFX-11 under the design conditions. According to this analysis, the maximum deflection in the Z-direction is $0.05104{\mu}m$ and the maximum temperature deviation is $0.7^{\circ}C$ when the temperature of the heating block unit is $175^{\circ}C$. It was also found that these values are structurally safe. The advantage of the developed multi-production forming machine is demonstrated to be in its offering of a proper voice test.

• Nuclear Engineering and Technology
• /
• 제53권2호
• /
• pp.386-392
• /
• 2021

MACCS (MELCOR Accident Consequence Code System), WinMACCS, and MelMACCS now facilitate a multi-unit consequence analysis. MACCS evaluates the consequences of an atmospheric release of radioactive gases and aerosols into the atmosphere and is most commonly used to perform probabilistic safety assessments (PSAs) and related consequence analyses for nuclear power plants (NPPs). WinMACCS is a user-friendly preprocessor for MACCS. MelMACCS extracts source-term information from a MELCOR plot file. The current development can combine an arbitrary number of source terms, representing simultaneous releases from a multi-unit facility, into a single consequence analysis. The development supports different release signatures, fission product inventories, and accident initiation times for each unit. The treatment is completely general except that the model is currently limited to collocated units. A major practical consideration for performing a multi-unit PSA is that a comprehensive treatment for more than two units may involve an intractable number of combinations of source terms. This paper proposes and evaluates an approach for reducing the number of calculations to be tractable, even for sites with eight or ten units. The approximation error introduced by the approach is acceptable and is considerably less than other errors and uncertainties inherent in a Level 3 PSA.