• Journal of the Korean Society of Safety
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• v.27 no.5
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• pp.229-234
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• 2012

Operator's action time is evaluated from MAAP4 analysis used in conventional probabilistic safety assessment(PSA) of a nuclear power plant. MAAP4 code which was developed for severe accident analysis is too conservative to perform a realistic PSA. A best-estimate code such as RELAP5/MOD3, MARS has been used to reduce the conservatism of thermal hydraulic analysis. In this study, operator's action time of core cooling recovery operation is evaluated by using the MARS code, which its Fussell-Vessely(F-V) value was evaluated as highly important in a small break loss of coolant(SBLOCA) event and loss of component cooling water(LOCCW) event in previous PSA. The main conclusions were elicited : (1) MARS analysis provides larger time window for operator's action time than MAAP4 analysis and gives the more realistic time window in PSA (2) Sufficient operator's action time can reduce human error probability and core damage frequency in PSA.

• Journal of the Korean Society of Safety
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• v.30 no.4
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• pp.174-180
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• 2015

This study developed a Safety Culture Impact Assessment Model (SCIAM) which consists of a safety culture assessment methodology and a safety culture impact quantification methodology. The SCIAM uses safety culture impact index (SCII) to monitor the status of safety culture of the NPPs periodically and it uses relative core damage frequency (RCDF) to present the impact of safety culture on the safety of the NPPs. As a result of applying SCIAM to the reference plant (Kori 3), the standard for the healthy safety culture of the reference plant is suggested. SCIAM might contribute to improve the safety of the NPPs (Nuclear Power Plants) by monitoring the status of safety culture periodically and presenting the standard of healthy safety culture.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.862-868
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• 1998

Korean Next Generation Reactor(KNGR) is a standardized evolutionary Advanced Light Water Reactor design under development Korea Power Company(KEPCO). It incorporates design enhncements such as active and passive advanced design features(ADFs) to increase the plant safety. A Preliminary level 1 Probabilistic Safety Assessment(PSA) has been performed for KNGR to examine the effect of these safety features. The preliminary PSA result shows that it meets the KNGR safety goal on core damage frequency(CDF). The result of this safety assessment shows that the four-train safety systems, and the ADFs such as Passive Secondary Cooling System (PSCS) contributes greatly to the reduction the CDF. Furthermore, several design changes are made or proposed for detailed review based on the PSA insights.

• Nuclear Engineering and Technology
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• v.44 no.5
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• pp.453-458
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• 2012

A major issue for all nuclear stakeholders is to keep the probability of circumstances that could lead to core damage as low as possible. In addition, for NPP, appropriate accident management provisions are to be implemented to limit the consequences associated with an accident. Development and application of L2 PSA is a structured way to demonstrate that such objectives are achieved. The paper presents the efforts recently done in Europe to harmonize some best-practices in that field, from research area to risk assessment. The Fukushima Daiichi accident reiterated the importance of these activities and the need to efficiently reinforce the NPP safety based on risk assessment conclusions. New perspectives in Europe are briefly presented.

• Nuclear Engineering and Technology
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• v.37 no.1
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• pp.109-117
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• 2005

Limiting conditions for operations (LCOs) are evaluated dynamically using the tool of system dynamics. The LCOs de-fine the allowed outage times (AOTs) and the actions to be taken if the repair cannot be completed within the AOT. System dynamics has been developed to analyze the dynamic reliability of a complicated system. System dynamics using Vensim software have been applied to LCOs assessment for an example system, the auxiliary feed water system of a reference nuclear power plant. Analysis results of both full power operation and shutdown operation have been compared for a measure of core damage frequency. The framework developed in this study has been shown to be very flexible in that it can be applied to assess LCOs quantitatively under any operational context of the TS in FSAR.

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1372-1386
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• 2018

Despite recent advances in multi-hazard analysis, the complexity and inherent nature of such problems make quantification of the landslide effect in a probabilistic safety assessment (PSA) of NPPs challenging. Therefore, in this paper, a practical approach was presented for performing an earthquake-induced landslide PSA for NPPs subject to seismic hazard. To demonstrate the effectiveness of the proposed approach, it was applied to Korean typical NPP in Korea as a numerical example. The assessment result revealed the quantitative probabilistic effects of peripheral slope failure and subsequent run-out effect on the risk of core damage frequency (CDF) of a NPP during the earthquake event. Parametric studies were conducted to demonstrate how parameters for slope, and physical relation between the slope and NPP, changed the CDF risk of the NPP. Finally, based on these results, the effective strategies were suggested to mitigate the CDF risk to the NPP resulting from the vulnerabilities inherent in adjacent slopes. The proposed approach can be expected to provide an effective framework for performing the earthquake-induced landslide PSA and decision support to increase NPP safety.

• Composites Research
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• v.27 no.2
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• pp.77-81
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• 2014

In this study, it was performed damage assessment and repair of small scale aircraft adopted on composite. This aircraft adopted the sandwich structure to skin of wing. This study aims to investigate the residual strength of sandwich composites with nomex honeycomb core and carbon fiber face sheets after the open hole damage by the experimental investigation. The 4-point bending tests were used to find the bending strength, and the open hole was applied to introduce the simulated damage on the specimen. The bending strength test results after open hole were compared with the results of no damaged specimen test. In addition, The damaged composite structure was repaired using external patch repair method after removing damaged area. After that, this study presents comparison results of the experimental investigation between the damaged and the repaired specimen. It was found that the bending strength of repaired specimen was recovered up to 95% of undamaged specimen.

• Earthquakes and Structures
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• v.21 no.5
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• pp.515-530
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• 2021

In a tall reinforced concrete (RC) core wall system subjected to strong ground motions, inelastic behavior near the base as well as mid-height of the wall is possible. Generally, the formation of plastic hinge in a core wall system may lead to extensive damage and significant repairing cost. A new configuration of core structures consisting of buckling restrained braced frames (BRBFs) and RC walls is an interesting idea in tall building seismic design. This concept can be used in the plan configuration of tall core wall systems. In this study, tall buildings with different configurations of combined core systems were designed and analyzed. Nonlinear time history analysis at severe earthquake level was performed and the results were compared for different configurations. The results demonstrate that using enough BRBFs can reduce the large curvature ductility demand at the base and mid-height of RC core wall systems and also can reduce the maximum inter-story drift ratio. For a better investigation of the structural behavior, the probabilistic approach can lead to in-depth insight. Therefore, incremental dynamic analysis (IDA) curves were calculated to assess the performance. Fragility curves at different limit states were then extracted and compared. Mean IDA curves demonstrate better behavior for a combined system, compared with conventional RC core wall systems. Collapse margin ratio for a RC core wall only system and RC core with enough BRBFs were almost 1.05 and 1.92 respectively. Therefore, it appears that using one RC core wall combined with enough BRBF core is an effective idea to achieve more confidence against tall building collapse and the results demonstrated the potential of the proposed system.

• Journal of Korean Society for Geospatial Information Science
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• v.24 no.1
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• pp.81-87
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• 2016

It is important to analyze the exact damage information for fast recovery when natural disasters cause damage on river-side facilities such as dams, bridges, embankments etc. In this study, we shows the method to effectively damage analysis plan using UAV(Unmanned aerial vehicle) images and accuracy assessment of it. The UAV images are captured on area near the river-side facilities and the core methodology for damage analysis are image matching and change detection algorithm. The result(point cloud) from image matching is to construct 3-dimensional data using by 2-dimensional images, it extracts damage areas by comparing the height values on same area with reference data. The results are tested absolute locational precision compared by post-processed aerial LiDAR data named reference data. The assessment analysis test shows our matching results 10-20 centimeter level precision if external orientation parameters are very accurate. This study shows suggested method is very useful for damage analysis in a large size structure like river-side facilities. But the complexity building can't apply this method, it need to the other method for damage analysis.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1414-1422
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• 2017

Remote response technology has advanced to the extent that a robot system, if properly designed and deployed, may greatly help respond to beyond-design-basis accidents at nuclear power plants. Particularly in the aftermath of the Fukushima accident, there is increasing interest in developing disaster robots that can be deployed in lieu of a human operator to the field to perform mitigating actions in the harsh environment caused by extreme natural hazards. The nuclear robotics team of the Korea Atomic Energy Research Institute (KAERI) is also endeavoring to construct disaster robots and, first of all, is interested in finding out to what extent safety benefits can be achieved by such a disaster robotic system. This paper discusses a new approach based on the probabilistic risk assessment (PRA) technique, which can be used to quantify safety benefits associated with disaster robots, along with a case study for seismic-induced station blackout condition. The results indicate that to avoid core damage in this special case a robot system with reliability > 0.65 is needed because otherwise core damage is inevitable. Therefore, considerable efforts are needed to improve the reliability of disaster robots, because without assurance of high reliability, remote response techniques will not be practically used.