• Nuclear Engineering and Technology
• /
• v.56 no.3
• /
• pp.819-823
• /
• 2024

The main criteria used in NEI 18-04 to define SSCs as risk-significant include (1) the SSC is required to keep all LBEs within the F-C target, and (2) the total frequency with the SSC failed exceeds 1% of the limit for at least one of the three cumulative risk metrics used for evaluating the integrated plant risk. The first one is a reasonable criterion in determining the risk significant SSCs. However, the second criterion may not be adequate to serve the purpose of determining the risk significance of SSCs. In the second criterion, the cumulative risk metric values representing the integrated plant risk (less the preventive and mitigative effects of the SSC being evaluated) are compared to a risk limit that represents a very small contribution to the overall integrated plant risk, which corresponds appropriately to the contributions from individual SSCs. The easiest approach to redefine the NEI 18-04 definition of risk-significant SSCs in relation to the integrated plant risk metrics is to compare the difference, between the risk metric value calculated with the SSC failed and the risk metric value calculated with the SSC credited, with 1% of the risk limit established for the integrated plant risk metrics.

• The Plant Pathology Journal
• /
• v.35 no.5
• /
• pp.459-472
• /
• 2019

The increasing variation in climatic conditions under climate change directly influences plant-microbe interactions. To account for as many variables as possible that may play critical roles in such interactions, the use of an integrated modeling approach is necessary. Here, we report for the first time a local impact assessment and adaptation study of future epidemics of kiwifruit bacterial blossom blight (KBB) in Jeonnam province, Korea, using an integrated modeling approach. This study included a series of models that integrated both the phenological responses of kiwifruit and the epidemiological responses of KBB to climatic factors with a 1 km resolution, under the RCP8.5 climate change scenario. Our results indicate that the area suitable for kiwifruit cultivation in Jeonnam province will increase and that the flowering date of kiwifruit will occur increasingly earlier, mainly due to the warming climate. Future epidemics of KBB during the predicted flowering periods were estimated using the Pss-KBB Risk Model over the predicted suitable cultivation regions, and we found location-specific, periodic outbreaks of KBB in the province through 2100. Here, we further suggest a potential, scientifically-informed, long-term adaptation strategy using a cultivar of kiwifruit with a different maturity period to relieve the pressures of future KBB risk. Our results clearly show one of the possible options for a local impact assessment and adaptation study using multiple models in an integrated way.

• Nuclear Engineering and Technology
• /
• v.39 no.2
• /
• pp.123-128
• /
• 2007

In order to apply a static fault-tree (FT) method to a system or a plant whose configuration changes dynamically, condition gates and a post processing method are used to effectively accommodate these changes. An operator's performance change, which can be caused by these configuration changes, should also be considered to assess the risk to a plant in a more realistic manner. This study aims to develop an integrated framework to accommodate various configuration changes and their effect on an operator’s performance by using the FT model. We applied a condition-based human reliability assessment (CBHRA) method to consider various conditions endured by an operator. That is, we integrated the CBHRA method with the conventional post processing method for modeling the system configuration changes. The effect of the condition monitoring systems installed in a plant is also considered. In this study, we show an example application of the integrated framework to a probabilistic safety assessment for the shutdown phase of a nuclear power plant.

• Nuclear Engineering and Technology
• /
• v.49 no.2
• /
• pp.426-433
• /
• 2017

This paper describes the development process, the innovative techniques used and insights gained from the latest integrated, full scope, multistate Level 2 PSA analysis conducted at the Leibstadt Nuclear Power Plant (KKL), Switzerland. KKL is a modern single-unit General Electric Boiling Water Reactor (BWR/6) with Mark III Containment, and a power output of $3600MW_{th}/1200MW_e$, the highest among the five operating reactors in Switzerland. A Level 2 Probabilistic Safety Assessment (PSA) analyses accident phenomena in nuclear power plants, identifies ways in which radioactive releases from plants can occur and estimates release pathways, magnitude and frequency. This paper attempts to give an overview of the advanced modeling techniques that have been developed and implemented for the recent KKL Level 2 PSA update, with the aim of systematizing the analysis and modeling processes, as well as complying with the relatively prescriptive Swiss requirements for PSA. The analysis provides significant insights into the absolute and relative importances of risk contributors and accident prevention and mitigation measures. Thanks to several newly developed techniques and an integrated approach, the KKL Level 2 PSA report exhibits a high degree of reviewability and maintainability, and transparently highlights the most important risk contributors to Large Early Release Frequency (LERF) with respect to initiating events, components, operator actions or seismic component failure probabilities (fragilities).

• Corrosion Science and Technology
• /
• v.11 no.2
• /
• pp.43-47
• /
• 2012

Recently, regulatory bodies quite often encourage to adopt risk-based inspection (RBI) and management programs because they can enhance safety simultaneously with deregulation in Korea. RBI is an integrated methodology that factors risk into inspection and maintenance decision making. This paper describes an example of how to use known risk assessment codes (API 580, API 581 BRD) to address such safety analysis requirements for risk management in the refining industry. Specifically, this paper reports the methodology and the results of implementation to the Crude Distillation Unit(CDU) plant of refinery units using the $KGS-RBI^{TM}$ program, developed by the Korea Gas Safety Corporation in reference of API Codes and ASME PC (Post Construction) with a suitable consideration of Korean situation. The results of the risk and reliability assessment using $KGS-RBI^{TM}$ program are useful in determining whether the detected defects are tolerable or required to be repaired. The subsequent decisions are to manage the future inspection, repair and maintenance planning in the risk reduction control.

• Nuclear Engineering and Technology
• /
• v.50 no.5
• /
• pp.627-638
• /
• 2018

This article describes an integrated Level 1-Level 2 probabilistic risk assessment (PRA) methodology to evaluate the radiological risk during postulated accident scenarios initiated during the decommissioning phase of a typical Mark I containment boiling water reactor. The fuel damage scenarios include those initiated while the reactor is permanently shut down, defueled, and the spent fuel is located into the spent fuel storage pool. This article focuses on the integrated Level 1-Level 2 PRA aspects of the analysis, from the beginning of the accident to the radiological release into the environment. The integrated Level 1-Level 2 decommissioning PRA uses event trees and fault trees that assess the accident progression until and after fuel damage. Detailed deterministic severe accident analyses are performed to support the fault tree/event tree development and to provide source term information for the various pieces of the Level 1-Level 2 model. Source terms information is collected from accidents occurring in both the reactor pressure vessel and the spent fuel pool, including simultaneous accidents. The Level 1-Level 2 PRA model evaluates the temporal and physical changes in plant conditions including consideration of major uncertainties. The goal of this article is to provide a methodology framework to perform a decommissioning Probabilistic Risk Assessment (PRA), and an application to a real case study is provided to show the use of the methodology. Results will be derived from the integrated Level 1-Level 2 decommissioning PSA event tree in terms of fuel damage frequency, large release frequency, and large early release frequency, including uncertainties.

• Fire Science and Engineering
• /
• v.17 no.4
• /
• pp.20-27
• /
• 2003

Hazard Analysis is one of the basic tasks to ensure the safety of chemical plants. However, it is an arduous, tedious, time-consuming work and requires multidisciplinary knowledge and demands considerable cognitive load from the analysts. To overcome these problems, there have been attempts to automate this work by utilizing computer technology, particularly in the area of knowledge-based technique. There is two methods in the risk assessment of Chemical plant; quantitative and qualitative risk assessment. Both of them have been applied respectively, but if the integrated method of quantitative and qualitative risk assessments is used, all of the advantage of two methods can be applied. It is difficult to carry out integrated risk management of chemical plant. Therefore, automated integration system of risk management is necessary. We developed S/W Automated System for Hazard Screening & Analysis(ASCA) and applied to practical plant. By applying ASCA to case study, we can get the information about relative ranks of equipments, variable deviation, and consequence of potential accident. In this study, we applied ASCA to the H.T.U(Hydrotreating Unit) of the process to produce aromatic material. We could know relative ranks of equipments, variable deviation of malfunction in storage tank, D-101, and consequence of potential accident using ASCA. If integrated risk management in the chemical plant is applied, we can develop the emergency plan and prevent the accident.

• Journal of the Computational Structural Engineering Institute of Korea
• /
• v.27 no.5
• /
• pp.437-450
• /
• 2014

In this study, we developed a methodology to determine the reference parameter for an aircraft impact induced risk assessment of nuclear power plant (NPP) using finite element impact analysis of containment building. The target structure used to develop the method of reference parameter selection is one of the typical Korean PWR type containment buildings. We composed a three-dimensional finite element model of the containment building. The concrete damaged plasticity model was used for the concrete material model. The steels in the tendon, rebar, and liner were modeled using the piecewise-linear stress-strain curves. To evaluate the correlations between structural response and each candidate parameter, we developed Riera's aircraft impact force-time history function with respect to the variation of the loading parameters, i.e., impact velocity and mass of the remaining fuel. For each force-time history, the type of aircraft is assumed to be a Boeing 767 model. The variation ranges of the impact velocity and remaining fuel percentage are 50 to 200m/s, and 30 to 90%, respectively. Four parameters, i.e., kinetic energy, total impulse, maximum impulse, and maximum force are proposed for candidates of the reference parameter. The wellness of the correlation between the reference parameter and structural responses was formulated using the coefficient of determination ($R^2$). From the results, we found that the maximum force showed the highest $R^2$ value in most responses in the materials. The simplicity and intuitiveness of the maximum force parameter are also remarkable compared to the other candidate parameters. Therefore, it can be concluded that the maximum force is the most proper candidate for the reference parameter to assess the aircraft impact induced risk of NPPs.

• Nuclear Engineering and Technology
• /
• v.51 no.5
• /
• pp.1307-1313
• /
• 2019

Recent analysis results with realistic assumptions provide the variability of operator allowable time for the initiation of aggressive cooldown under small break loss of coolant accident or steam generator tube rupture with total failure of high pressure safety injection. We investigated how plant risk may vary depending on the variability of operators' failure probability of timely initiation of aggressive cooldown. Using a probabilistic safety assessment model of a nuclear power plant, we showed that plant risks had a linear relation with the failure probability of aggressive cooldown and could be reduced by up to 10% as aggressive cooldown is more reliably performed. For individual accident management, we found that core damage potential could be gradually reduced by up to 40.49% and 63.84% after a small break loss of coolant accident or a steam generator tube rupture, respectively. Based on the importance of timely initiation of aggressive cooldown by main control room operators within the success criteria, implications for improvement of emergency operating procedures are discussed. We recommend conducting further detailed analyses of aggressive cooldown, commensurate with its importance in reducing risks in nuclear power plants.

• Journal of Environmental Science International
• /
• v.24 no.7
• /
• pp.927-937
• /
• 2015

This study was performed to identify and quantify the asbestos fibers released from two types of asbestos-cement slate roofs. One is a plant roof installed in 1987 which contained 15% chrysotile, and the other is a residential roof installed before 1983 which contained 12% chrysotile. The concentrations of asbestos fibers in air surrounding asbestos-cement slate roofs and in the falling water harvested from the same roofs on rainy days ranged from 0.0012 to 0.0018 f/mL and from 1,764 f/L to 10,584 f/L, respectively. The concentration of inorganic fibers in the soil around asbestos-cement slate roofs was from 217 to 348 f/g. With the above results, the excess lifetime cancer risk (ELCR) for the risk assessment of the asbestos fibers released from asbestos-cement slate based on US EPA IRIS (Integrated risk information system) model is within 5.5E-06 ~ 6.5E-06 levels which indicates that the levels do not exceed "the acceptable risk(1.0E-05)" recommended by WHO. The asbestos concentration in air, drained rainfall and soil around the plant slate roof was higher than that around residential slate roof, but the excess lifetime cancer risk (ELCR) from residential slate was higher than that from plant slate. This suggested that the enclose and encapsulation of residential roofs have priority in removal policy to minimize the exposure risk.