• Nuclear Engineering and Technology
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• v.53 no.2
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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.

• Journal of the Korea Safety Management & Science
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• v.16 no.3
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• pp.195-200
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• 2014

Occupational fatal injury rate per 10,000 population of Korea is still higher among the OECD member countries. To prevent fatal injuries, the causes of accidents including human error should be analyzed and then appropriate countermeasures should be established. There was an severe converter furnace accident resulting in five people death by chocking in 2013. Although the accident type of the furnace accident was suffocation, many safety problems were included before reaching the death of suffocation. If the safety problems are reviewed throughly, the alternative measures based on the review would be very useful in preventing similar accidents. In this study, we investigated the converter furnace accident by using human error analysis and accident scenario analysis. As a result, it was found that the accident was caused by some human errors, inappropriate task sequence and lack of control in coordinating work by several subordinating companies. From the review of this case, the followings are suggested: First, systematic human error analysis should be included in the investigation of fatal injury accidents. Second, multi man-machine accident scenario analyis is useful in most of coordinating work. Third, the more provision of information on system state will lessen human errors. Fourth, the coordinating control in safety should be performed in the work conducting by several different companies.

• Nuclear Engineering and Technology
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• v.16 no.3
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• pp.141-154
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• 1984

An analysis is presented of key phenomena and scenario which imply some general trends for beyond design-basis-accident in Kori-1 PWR dry containment. The study covers a wide range of severe accident sequences initiated by small break LOCA. The MARCH computer code, with KAERI modifications was used in this analysis. The major emphasis of the paper are two folds, 1) the phenomenologic understanding of severe accident and 2) a study of H2 combustion and debris/ water interactions in a specific small break LOCA for Kori-1 plant. The sensitivity studies for the specific plant data and thermal interaction modelings used in the SASA were performed. The results show that if hydrogen burning does occur at low concentration, the resulting peak pressure does not exceed the design value, while the lower concentration assumption results in repeated burning due to the continuing H$_2$ generation. For debris/water interaction, the particle size has no effect on the magnitude of peak pressure for the amount of water assumed to be in the reactor cavity. But, the occurrence of peak pressure is considerably delayed in case of using the dryout correlation. The peak containment pressure predicted from the hydrogen combustion and steam pressure spite during full core meltdown scenario does not present a severe threat to the containment integrity.

• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.197-201
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• 2001

A risk analysis consists of a triplet, , where Si is the scenario identification; Pi is the probability of each scenario; and Xi is the consequences of each scenario. A new computing framework, OMAM (ORIGEN-MAAP4-MMCS), has been developed and applied for assessing the risk of a reference plant as well as radiation source terms using the concept of risk triplet. The result of this study using the OMAM framework presented in this paper, can contribute to producing domestic nuclear power plant's risk data base as well as to establishing severe accident management plans.

• Nuclear Engineering and Technology
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• v.36 no.1
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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

• Nuclear Engineering and Technology
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• v.28 no.3
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• pp.290-298
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• 1996

Direct Containment Heating, DCH, is supposed to occur during a core melt-down accident if the primary system pressure is still high at the time of vessel breach in a Nuclear Power Plant (NPP). In this case, DCH is considered to be one of very important severe phenomena during postulated severe accident scenario because of the fast heat transfer rate to atmosphere and the sharp pressure increase in a containment. To reduce the effect of this DCH phenomena, the capture volume wes designed at Ulchin NPP units 3 and 4. But, the effect of this has not been studied extensively. This work consists of experimental and numerical analyses of the effects of capture volume in the cavity on DCH phenomena. The experimental model is a 1/30 scaled-down model of Ulchin NPP units 3 and 4. We used three types of capture volumes to investigate the effect of size. Numerical analysis using CONTAIN 1.2 is performed with the correlation for the dispersed fraction of molten corium from the cavity into the containment derived from the experimental data to examine the effect of capture volume on DCH phenomena in full scale of Ulchin NPP units 3 and 4.

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1343-1354
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• 2022

Experiments were performed to measure the critical heat flux (CHF) on a vertical surface abutting a coarse packed bed of spherical particles. This geometry is representative of a CANDU reactor calandria tubesheet facing the end shield cavity during the in-vessel retention (IVR) phase of a severe accident. Deionized light water was used as the working fluid. Low carbon steel shielding balls with diameters ranging from 6.4 to 12.7 mm were used, allowing for the development of an empirical correlation of CHF as a function of shielding ball diameter. Previously published data is used to develop a more comprehensive empirical correlation accounting for the impacts of both shielding ball diameter and heating surface height. Tests using borosilicate shielding balls demonstrated that the dependence of CHF on shielding ball thermal conductivity is insignificant. The deposition of iron oxide particles transported from shielding balls to the heating surface is verified to increase CHF non-trivially. The results presented in this paper improve the state of the knowledge base permitting quantitative prediction of CHF in the CANDU end shield, refining our ability to assess the feasibility of IVR. The findings clarify the mechanisms governing CHF in this scenario, permitting identification of potential future research directions.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.866-872
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• 2024

Design Extension Conditions (DEC)-A assessments of the operating nuclear power plants are generally considered for the purpose of getting additional safety demonstrations of their capability to undergo conditions that are generally more severe than DBAs by features implemented in the design and accident management measures. The pursued methodology is generally based upon Best Estimate approaches aiming at verifying that the safety limits in terms of integrity of the barriers against eventual large or early releases of radioactive material are fulfilled. These aspects are nowadays being experimentally and analytically addressed within the OECD/NEA experimental projects like the ATLAS and PKL series where a set of DEC-A experiments are considered. In this paper, experiments related to SBLOCA at the vessel upper head of the pressurized vessel of ATLAS and PKL are analytically assessed using the CATHARE code. These experiments includes issues related to common cause failure of the safety injection system and operator actions for preventing core excessive overheating. It is shown that, on the one hand, the safety features embedded in the design together with the operator actions are capable to prevent the progression towards a severe accident state and on the other hand, the code prediction capabilities for such scenario are generally good but still to be enhanced.

• Proceedings of the KSR Conference
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• 2005.11a
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• pp.742-747
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• 2005

Electric railroad transformer of a supply of Operation power of DC electric cars is intense fluctuation of load and flows the only big short-circuit current as a accident of the power system. it is a peculiarity more severe than general power transformer. Consequently, researches the properties about the rectifier mold transformer of DC substation and applies with data of safety of the electric railroad transformer. This paper analyzed a failure mode, the accident occurrence scenario and the be latent dangerous unit against the rectifier mold transformer of DC railway system.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.1
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• pp.36-43
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• 2023

Lots of researches have been conducted on in-vessel retention (IVR) to prevent or mitigate severe accident in nuclear power plants. Various methodologies were proposed and the external reactor vessel cooling was selected as a part of promising IVR strategy. In this study, the strategy is strengthened by enhancing the natural circulation performance through the adoption of insulation in the reactor cavity. A thermal analysis was carried out based on an assumed accident scenario and its results were used as boundary conditions for subsequent seven flow analysis cases. By comparing the natural circulation performance, effects of annular gaps and insulation shapes on the mass flow rate and flow velocity were quantified. The improvement in cooling performance can be reflected in actual design via detailed assessment.