• Korean Journal of Hazardous Materials
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• v.3 no.1
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• pp.52-58
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• 2015

Since the HF release in 2012 in Korea, it became one of the most significant to evaluate consequence to the vicinity of industry facilities handling hazardous materials. BTX plant is selected to assess off-site risk to check whether the facility satisfies the Chemical Control Law by Korea Government. Accident scenarios were listed using process safety information. The scenarios having effect to the off-site were selected and assessed further according to guideline provided by Korea government. Worst case and alternative scenarios including other interested scenarios were evaluated using ALOHA. Each evaluated scenario was assessed further considering countermeasures. The results showed that the facility handling chloric acid is safe enough and needed no further protections at the moment.

• Korean System Dynamics Review
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• v.16 no.1
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• pp.53-74
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• 2015

The purpose of this paper is to simulate public trust on nuclear regulation policy. The first of all, public trust variables and the model were developed and analysed by system dynamic method. The model are consisted of the operator safety culture level, regulatory competence levels, the public satisfaction and public trust level. The scenario is made up three type which base scenario, the system operator's safety culture level and accident event level. First. the simulation results of standard scenario shows that rapidly declining public satisfaction and trust level of the national safety after Japan's nuclear accident in November 2011. Second, operator safety culture level and simulated divided into three levels. The results showed that a greater impact on the public satisfaction if bad than good case. Finally, the size of the accident was simulated divided into three levels levels(no accident, medium, serious accidents). the results showed a weak effect against the regulatory capacity and safety performance levels but showed a significant impact on public satisfaction and confidence level.

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1246-1254
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• 2018

The importance of performing Level 3 probabilistic safety assessments (PSA) along with a general interest in assessing multi-unit risk has been sharply increasing after the Fukushima Daiichi nuclear power plant (NPP) accident. However, relatively few studies on multi-unit Level 3 PSA have been performed to date, reflecting limited scenarios of multi-unit accidents with higher priority. The major difficulty to carry out a multi-unit Level 3 PSA lies in the exponentially increasing number of multi-unit accident combinations, as different source terms can be released from each NPP unit; indeed, building consequence models for the astronomical number of accident scenarios is simply impractical. In this study, a new approach has been developed that employs the look-up table method to cover every multi-unit accident scenario. Consequence results for each scenario can be found on the table, established with a practical amount of effort, and can be matched to the frequency of the scenario. Preliminary application to a six-unit NPP site was carried out, where it was found that the difference between full-coverage and cut-off cases could be considerably high and therefore influence the total risk. Additional studies should be performed to fine tune the details and overcome the limitations of the approach.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.999-1008
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• 2023

Deformed fuel rods can cause a partial blockage of the flow area in a subchannel. Such flow blockage will influence the core coolant flow and further the core heat transfer during the reflooding phase and subsequent severe accidents. Nevertheless, most of the system analysis codes simulate the accident process based on the assumed flow blockage ratio, resulting in inconsistencies between simulated results and actual conditions. This paper aims to study the influence of flow blockage in severe accident scenario of the CAP1400 reactor. First, the flow blockage model of ISAA code is improved based on the FRTMB module. Then, the ISAA-FRTMB coupling system is adopted to model and calculate the QUENCH-LOCA-0 experiment. The correctness and validity of the flow blockage model are verified by comparing the peak cladding temperature. Finally, the DVI Line-SBLOCA accident is induced to analyze the influence of flow blockage on subsequent CAP1400 reactor core heat transfer and core degradation. From the results of the DVI Line-SBLOCA accident analysis, it can be concluded that the blockage ratio is in the range of 40%-60%, and the position of severe blockage is the same as that of cladding rupture. The blockage reduces the circulation area of the core coolant, which in turn impacts the heat exchange between the core and the coolant, leading to the early failure and collapse of some core assemblies and accelerating the core degradation process.

• Journal of the Korean Society of Systems Engineering
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• v.19 no.2
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• pp.32-45
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• 2023

Implementing Severe Accident Management (SAM) strategies is crucial for enhancing a nuclear power plant's resilience and safety against severe accidents conditions represented in the analysis of Station Blackout (SBO) event. Among these critical approaches, the In-Vessel Retention (IVR) through External Reactor Vessel Cooling (IVR-ERVC) strategy plays a key role in preventing vessel failure. This work is designed to evaluate the efficacy of the IVR strategy for a high-power density reactor APR1400. The APR1400's plant is represented and simulated under steady-state and transient conditions for a station blackout (SBO) accident scenario using the computer code, ASYST. The APR1400's thermal-hydraulic response is analyzed to assess its performance as it progresses toward a severe accident scenario during an extended SBO. The effectiveness of emergency operating procedures (EOPs) and severe accident management guidelines (SAMGs) are systematically examined to assess their ability to mitigate the accident. A group of associated key phenomena selected based on Phenomenon Identification and Ranking Tables (PIRT) and uncertain parameters are identified accordingly and then propagated within DAKOTA Uncertainty Quantification (UQ) framework until a statistically representative sample is obtained and hence determine the uncertainty bands of key system parameters. The Systems Engineering methodology is applied to direct the progression of work, ensuring systematic and efficient execution.

• Journal of Auto-vehicle Safety Association
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• v.10 no.2
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• pp.20-28
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• 2018

Currently, sharp increase of car is on the rise as a serious social problem due to loss of lives from car accident and environmental pollution. There is a study on ITS (Intelligent Transportation System) to seek coping measures. As for the commercialization of ITS, we aim for occupancy of world market through ASV (Advanced Safety Vehicle) related system development and international standardization. However, the domestic environment is very insufficient. Core factor technologies of ITS are Adaptive Cruise Control, Lane Keeping Assist System, Forward Collision Warning System, AEB (Autonomous Emergency Braking) system etc. These technologies are applied to cars to support driving of a driver. AEB system is stop the car automatically based on the result decided by the relative speed and distance with obstacle detected through sensor attached on car rather than depending on the driver. The purpose of AEB system is to measure the distance and speed of car and to prevent accident. Thus, AEB will be a system useful for prevention of accident by decreasing car accident along with the development of automobile technology. This study suggests a scenario to suggest a test evaluation method that accords with domestic environment and active response of international standard regarding the test evaluation method of AEB. Also, by setting the goal with function for distance, it suggests theoretic model according to the result. And the study aims to verify the theoretic evaluation standard per proposed scenario using car which is installed with AEB device through field car driving test on test road. It will be useful to utilize the suggested scenario and theoretical model when conducting AEB test evaluation.

• Journal of the Korean Society of Marine Environment & Safety
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• v.24 no.7
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• pp.863-869
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• 2018

Due to recent changes in the maritime traffic environment, naval warship accidents are constantly occurring. Especially in 2017, serious loss of life was caused by a US navy destroyer accident. The purpose of this study is to analyze the characteristics of naval warship accident cases and construct an accident scenario by using naval training materials, adjudication of naval warship accidents and US navy destroyer accident reports. Based on the surveyed data, the status of accidents was identified and cases were analyzed. We reproduced 17 accident cases in accordance with accident reproduction procedure and constructed naval warship accident scenarios. As a result of analyzing the CPA, TCPA and PARK model for risk, reproducing 17 naval ship accident cases, collision risk increased on average 5-6 minutes before an accident. The result of this study represents basic data for naval and simulation education materials, contributing to the prevention of marine accidents.

• Korean Chemical Engineering Research
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• v.54 no.6
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• pp.781-785
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• 2016

Chloric acid is a toxic chemical and the risk of facility handling chloric acid was assessed from the list of accident scenario to provide countermeasure to keep the vicinity safe. Accident scenarios were listed by using MSDS and process safety information. The scenarios having effect to the off-site were selected and assessed further according to guideline provided by Korea government. Worst case and alternative scenarios including other interested scenarios were evaluated using ALOHA. Each evaluated scenario was assessed further considering countermeasures. The results showed that the facility handling chloric acid is safe enough and needed no further protections at the moment.

• Journal of Radiation Protection and Research
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• v.48 no.1
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• pp.52-57
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• 2023

Background: Korea Atomic Energy Research Institute (KAERI) operates several nuclear research facilities licensed by Nuclear Safety and Security Commission (NSSC). The emergency preparedness requirements, GSR Part 7, by International Atomic Energy Agency (IAEA) request protection strategy based on the hazard assessment that is not applied in Korea. Materials and Methods: In developing the protection strategy, it is important to consider an accident scenario and its consequence. KAERI has tried the hazard assessment based on a hypothesis accident scenario for the major nuclear facilities. During the assessment, the safety analysis report of the related facilities was reviewed, the simulation using MELCOR, MACCS2 code was implemented based on a considered accident scenario of each facility, and the international guidance was considered. Results and Discussion: The results of the optimized protective actions were 300 m evacuation and 800 m sheltering for the High-Flux Advanced Neutron Application Reactor (HANARO), the evacuation to radius 50 m, the sheltering 400 m for post-irradiation examination facility (PIEF), 100 m evacuation or sheltering for HANARO fuel fabrication plant (HFFP) facility. Conclusion: The results of the optimized protective actions and its distances for the KAERI facilities for the maximum postulated accidents were considered in establishing the emergency plan and procedures and implementing an emergency exercise for the KAERI facilities.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2859-2865
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• 2021

One of the important issues raised from the Fukushima-Daiichi accident is the safety of multi-unit sites when simultaneous accidents occur at the site and recently a multi-unit PSA methodology is being developed worldwide. Since all operation modes of the plant should be considered in the multi-unit PSA, the accident analysis needs to be performed for shutdown operation modes, too. In this study, a station blackout during the mid-loop operation is selected as a reference scenario. The overall accident progression for the mid-loop operation is slower than that for the full-power operation because the residual heat per mass of coolant is about 6 times lower than that in the mid-loop scenario. Though the fractions of Cs released from the core to the RCS in both operation modes are almost the same, the amount of Cs delivered to the containment atmosphere is quite different due to the chemisorption in the RCS. While 45.5% of the initial inventory is chemisorbed on the RCS surfaces during the full-power operation, only 2.2% during the mid-loop operation. The containment remains intact during the mid-loop operation, though 83.9% of Cs is delivered to the containment.