• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.469-474
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• 1997

To provide tile information ell severe accident progression is very important for advanced or new type of nuclear power plant (NPP) design. A parametric study, therefore was performed to investigate the effect of thermal hydraulic design parameters ell severe accident progression of pressurized water reactors (PWRs), Nine parameters, which are considered important in NPP design or severe accident progression, were selected among the various thermal hydraulic design parameters. The backpropagation neural network (BPN) was used to determine parameters, which might more strongly affect the severe accident progression, among mile parameters. For training. different input patterns were generated by the latin hypercube sampling (LHS) technique and then different target patterns that contain core uncovery time and vessel failure time were obtained for Young Gwang Nuclear (YGN) Units 3&4 using modular accident analysis program (MAAP) 3.0B code. Three different severe accident scenarios, such as two loss of coolant accidents (LOCAs) and station blackout(SBO), were considered in this analysis. Results indicated that design parameters related to refueling water storage tank (RWST), accumulator and steam generator (S/G) have more dominant effects on the progression of severe accidents investigated, compared to tile other six parameters.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.484-497
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• 2021

This paper proposes an extended time uncertainty analysis approach in Level 2 human reliability analysis (HRA) considering severe accident management (SAM) strategies. The method is a time-based model that classifies two time distribution functions-time required and time available-to calculate human failure probabilities from delayed action when implementing SAM strategies. The time required function can be obtained by the combination of four time factors: 1) time for diagnosis and decision by the technical support center (TSC) for a given strategy, 2) time for strategy implementation mainly by the local emergency response organization (ERO), 3) time to verify the effectiveness of the strategy and 4) time for portable equipment transport and installation. This function can vary depending on the given scenario and includes a summation of lognormal distributions and a choice regarding shifting the distribution. The time available function can be obtained via thermal-hydraulic code simulation (MAAP 5.03). The proposed approach was applied to assess SAM strategies that use portable equipment and safety depressurization system valves in a total loss of component cooling water event that could cause reactor vessel failure. The results from the proposed method are more realistic (i.e., not conservative) than other existing methods in evaluating SAM strategies involving the use of portable equipment.

• Nuclear Engineering and Technology
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• 제56권7호
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• pp.2859-2874
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• 2024

This paper presents the finite element deformation and failure simulation of a typical Korean high-power reactor vessel under a severe accident characterized by large break loss of coolant (LBLOCA) with in-vessel retention of molten corium through external reactor vessel cooling (IVR-ERVC) conditions. Temperature distributions calculated using Modular Accident Analysis Program Version 5 (MAAP5) as thermal boundary conditions were used, and ABAQUS thermal and structural analyses were performed. After full ablation, the temperature of the inner surface in the thinnest section remained high (920 ℃), but the stress remained relatively low (less than 6 MPa). At the outer surface, the stress was as high as 250 MPa; however, the resulting plastic strain was small owing to the low temperature of 200 ℃. Variations in stress, inelastic strain, and temperature with time in the thinnest section suggest that the plastic and creep strains are saturated owing to stress relaxation, resulting in low cumulative damage. Thus, the lower head of the vessel can maintain its structural integrity under LBLOCA with IVR-ERVC conditions. The sensitivity analysis of internal pressure indicates the occurrence of failure in the thinnest section at an internal pressure >9.6 MPa via local necking followed by failure due to high stresses.

• 한국안전학회지
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• 제27권5호
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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.

• 한국신뢰성학회:학술대회논문집
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• 한국신뢰성학회 2000년도 춘계학술대회 발표논문집
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• pp.239-246
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• 2000

이 논문은 중대사고가 발생하였을때 운전원에 의하여 사고관리 방안을 수행하는 경우 그 실현성(Feasibility)을, 평가하는데 사용할 수 있는 새로운 시간의존적 신뢰도 분석방법을 제시하였다. 이 방법은 성능요구 (Performance Requirement)와 성능성취 (Performance Achievement)의 상관관계의 개념을 이용하는 신뢰도물리(Reliability Physics)와 모든 시간의존적 사고경위를 도출하는 동적사건수목 생성방법에 기초하고있다. 신뢰도물리는 성능요구변수와 성능성취변수의 비교를 이용한 신뢰도분석방법인 반면 동적사건수목 생성방법은 바람직한 해를 얻을 때까지 모든 가능한 사고경위를 도출해 내는 방법이다. 이 방법론을 정전사고시 참조원전의 공동에 비상화재시스템을 이용하여 물을 공급하는 공동범람사고관리 방안에 적용시켰다. Latin Hypercube Sampling 방법은 성능요구변수의 불확실성을 평가하는데 사용되었다. 제시된 방법론은 사고시 필요한 운전원의 방안수행 성공가능성을 평가하는데 사용될 수 있을 뿐만 아니라 궁극적으로 사고관리 절차서 개발에 도움이 될 수 있음을 보여주었다.

• Nuclear Engineering and Technology
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• 제47권1호
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• pp.74-84
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• 2015

If a transient occurs in a nuclear power plant (NPP), operators will try to protect the NPP by estimating the kind of abnormality and mitigating it based on recommended procedures. Similarly, operators take actions based on severe accident management guidelines when there is the possibility of a severe accident occurrence in an NPP. In any such situation, information about the occurrence time of severe accident-related events can be very important to operators to set up severe accident management strategies. Therefore, support systems that can quickly provide this kind of information will be very useful when operators try to manage severe accidents. In this research, the occurrence times of several events that could happen during a severe accident were predicted using support vector machines with short time variations of plant status variables inputs. For the preliminary step, the break location and size of a loss of coolant accident (LOCA) were identified. Training and testing data sets were obtained using the MAAP5 code. The results show that the proposed algorithm can correctly classify the break location of the LOCA and can estimate the break size of the LOCA very accurately. In addition, the occurrence times of severe accident major events were predicted under various severe accident paths, with reasonable error. With these results, it is expected that it will be possible to apply the proposed algorithm to real NPPs because the algorithm uses only the early phase data after the reactor SCRAM, which can be obtained accurately for accident simulations.

• Nuclear Engineering and Technology
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• 제36권1호
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• pp.24-35
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• 2004

In order to analyze the hydrogen distribution during a severe accident in the APR1400 containment, GASFLOW II was used. For the APR1400 NPP, a hydrogen mitigation system is considered from the design stage, but a fully time-dependent, three-dimensional analysis has not been performed yet. In this study GASFLOW code II is used for the three-dimensional analysis. The first step to analysis involving hydrogen behavior in a full containment with the GASLOW code is to generate a realistic geometry model, which includes nodalization and modeling of the internal structures such as walls, ceilings and equipment. Geometry modeling of the APR1400 is conducted using GUI program by overlapping the containment cut drawings in a graphical file format on the mesh view. The total number of mesh cells generated is 49,476. And the calculated free volume of the APR1400 containment by GASFLOW is almost the same as the value from the GOTHIC modeling. A hypothetical SB-LOCA scenario beyond design base accident was selected to analyze the hydrogen behavior with the hydrogen mitigation system. The source of hydrogen and steam for the GASFLOW II analysis is obtained from a MAAP calculation. Combustion pressure and temperature load possibilities within the compartments used in the GOTHIC analysis are studied based on the Sigma-Lambda criteria. Finally the effectiveness of HMS installed in the APR1400 containment is evaluated from the point of severe accident management

• 한국대기환경학회지
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• 제33권3호
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• pp.205-216
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• 2017

Carbonaceous aerosols such as the equivalent black carbon (eBC), the elemental carbon (EC) and the organic carbon (OC) were monitored at the Seoul Olympic Park site ($37.521^{\circ}N$, $127.124^{\circ}E$) during the KORUS-AQ 2016 campaign using a Multi Angle Absorption Photometer (MAAP) and an OCEC Analyzer. Averaged mass concentrations of eBC, EC and OC were presented as $2.46{\pm}1.52{\mu}g/m^3$, $1.01{\pm}0.60{\mu}g/m^3$ and $4.85{\pm}2.60{\mu}g/m^3$, respectively. OC/EC ratio and mass absorption cross-section (MAC) of light absorbing aerosols were calculated as 2.32 and $14.8m^2/g$, respectively. Diesel OC concentrations were estimated from a source profile of diesel vehicles as well. eBC mass concentrations measured from May $26^{th}$ to May $27^{th}$, 2016 showed 40% higher than averaged eBC mass concentrations during campaign period. $PM_{2.5}$ concentrations measured in this period were also higher than average $PM_{2.5}$ concentrations. High eBC concentrations were observed from May $29^{th}$ to May $31^{st}$, 2016 and from June $9^{th}$ to June $11^{th}$, 2016, possibly due to morning rush hour and the effect of temperature inversion at night. Diurnal variations of eBC, EC and Diesel OC showed a typical pattern of metropolitan area. In the weekend, however, diurnal variations of eBC, EC and Diesel OC mass concentrations were different from those measured in the weekday. It is expected that this study can help to understand the relationship between carbonaceous aerosols in a metropolitan area.