• Journal of the Earthquake Engineering Society of Korea
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• v.17 no.2
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• pp.53-59
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• 2013

Several researches have been studied to enhance the seismic performance of nuclear power plants (NPPs) by application of seismic isolation. If a seismic base isolation system is applied to NPPs, seismic performance of nuclear power plants should be reevaluated considering the soil-structure interaction effect. The seismic fragility analysis method has been used as a quantitative seismic safety evaluation method for the NPP structures and equipment. In this study, the seismic performance of an isolated NPP is evaluated by seismic fragility curves considering the soil-structure interaction effect. The designed seismic isolation is introduced to a containment building of Shin-Kori NPP which is KSNP (Korean Standard Nuclear Power Plant), to improve its seismic performance. The seismic analysis is performed considering the soil-structure interaction effect by using the linearized model of seismic isolation with SASSI (System for Analysis of Soil-Structure Interaction) program. Finally, the seismic fragility is evaluated based on soil-isolation-structure interaction analysis results.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.644-654
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• 2024

After the Tohoku earthquake and tsunami (Japan, 2011), regulatory efforts to mitigate external hazards have increased both the safety requirements and the total capital cost of nuclear power plants (NPPs). In these circumstances, identifying not only disaster robustness but also cost-effective capacity setting of NPPs has become one of the most important tasks for the nuclear power industry. A few studies have been performed to relocate the seismic capacity of NPPs, yet the effects of multiple hazards have not been accounted for in NPP capacity optimization. The major challenges in extending this problem to the multihazard dimension are (1) the high computational costs for both multihazard risk quantification and system-level optimization and (2) the lack of capital cost databases of NPPs. To resolve these issues, this paper proposes an effective method that identifies the optimal multihazard capacity of NPPs using a multi-objective genetic algorithm and the two-stage direct quantification of fault trees using Monte Carlo simulation method, called the two-stage DQFM. Also, a capacity-based indirect capital cost measure is proposed. Such a proposed method enables NPP to achieve safety and cost-effectiveness against multi-hazard simultaneously within the computationally efficient platform. The proposed multihazard capacity optimization framework is demonstrated and tested with an earthquake-tsunami example.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.1004-1012
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• 2021

High density polyethylene (HDPE) pipe has many advantages over metallic pipe, and has been used in non-safety related application for years in some nuclear power plants (NPPs). Recently, HDPE pipe was introduced into safety related applications. The main difference between safety-related and non-safety-related pipes in NPPs is the design method of extra loadings such as gravity, temperature, and earthquake. In this paper, the mechanical behavior of HDPE pipe under various loads in pipe gallery was studied by finite element analysis (FEA). Stress concentrations were found at the fusion regions on inner surface of mitered elbows of HDPE pipe system. The effects of various factors were analyzed, and the influence of various loads on the damage of HDPE pipe system were evaluated. The results of this paper provide a reference for the design of nuclear safety-related Class 3 HDPE pipe. In addition, as the HDPE pipes analyzed in this paper were suspended in pipe gallery, it can also serve as a supplementary reference for current ASME standard on Class 3 HDPE pipe, which only covers the application for buried pipe application.

• Journal of the Ergonomics Society of Korea
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• v.23 no.2
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• pp.13-24
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• 2004

Human error has brought about accidents more than 50% in system of a large size and complicated expecially in nuclear power plants(NPPs). The technology of Man Machine Interface(MMI) has been changed to the digitalized controls employing computer-based technology. According to this trend. the human factors guidelines are becoming main issue for reliable supports to digitalized information displays. However. the existing human factors guidelines is not enough for advanced information display on NPPs. The purpose of this research is to develop the reliable design and evaluation guidelines for advanced information display in main control room (MCR) of NPPs. In this study. the various general human factors guidelines concerning information display on CRT are integrated on data base management system. unified based on the integration rules. and applied in computer based procedures. The use of the integrated guidelines are expected to evaluate the existing information display on MCR in NPPs from the human factors point of view.

• Corrosion Science and Technology
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• v.14 no.6
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• pp.325-330
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• 2015

Pipe wall-thinning by flow-accelerated corrosion (FAC) is a significant and costly damage of secondary system piping in nuclear power plants (NPPs). All NPPs have their management programs to ensure pipe integrity from wall-thinning. This study analyzed the pipe wall-thinning caused by changing the amine, which is used for adjusting the water chemistry in the secondary system of NPPs. The pH change was analyzed according to the addition of amine. Then, the wear rate calculated in two different amines was compared at the steam cycle in NPPs. As a result, increasing the pH at operating temperature (Hot pH) can reduce the rate of FAC damage significantly. Wall-thinning is affected by amine characteristics depending on temperature and quality of water.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3043-3050
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• 2022

In Korea, all nuclear power plants (NPPs) participate in annual performance tests including in vivo measurements using the FastScan, a stand type whole body counter (WBC), manufactured by Canberra. In 2018, all Korean NPPs satisfied the testing criterion, the root mean square error (RMSE) ≤ 0.25, for the whole body configuration, but three NPPs which participated in an additional lung configuration test in the fission and activation product category did not meet the criterion. Due to the low resolution of the FastScan NaI(Tl) detectors, the conventional peak analysis (PA) method of the FastScan did not show sufficient performance to meet the criterion in the presence of interfering radioisotopes (RIs), ¹³⁴Cs and ¹³⁷Cs. In this study, we developed an artificial neural network (ANN) to improve the performance of the FastScan in the lung configuration. All of the RMSE values derived by the ANN satisfied the criterion, even though the photopeaks of ¹³⁴Cs and ¹³⁷Cs interfered with those of the analytes or the analyte photopeaks were located in a low-energy region below 300 keV. Since the ANN performed better than the PA method, it would be expected to be a promising approach to improve the accuracy and precision of in vivo FastScan measurement for the lung configuration.

• Fire Science and Engineering
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• v.34 no.3
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• pp.134-140
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• 2020

Nuclear power plants (NPPs) in Korea are required to be maintained using a defense in-depth approach to prevent leakage of radioactive substances outside the plant and allow safe shutdown in the event of a fire. Periodic testing must be conducted to ensure that the fire protection facilities perform as required by the laws for various nuclear reactor types. In June 2017, for the first time in Korea, a nuclear plant, Kori Unit 1, was permanently shut down. It was prepared for decommissioning in accordance with the fire protection regulations imposed by the regulatory body. However, a standard protocol is necessary for systematically establishing the fire protection program for decommissioning of NPPs in the future. Therefore, the nuclear legal systems of countries with many operating nuclear power plants, such as the United States, Japan, Canada, and various European countries, were reviewed and guidelines for establishing a fire protection program for decommissioning NPPs was suggested; the fire protection requirements stated by Reg Guide 1.191 (Decommissioning fire protection program for NPPs during decommissioning and permanent shutdown) were used as a model. Suggestions for establishing legal regulations to optimize fire protection programs and secure basic technology for decommissioning NPPs were also made.

• Nuclear Engineering and Technology
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• v.52 no.11
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• pp.2652-2659
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• 2020

The methyl iodide (CH₃I) removal performance of tri-ethylene-di-amine impregnated activated carbon (TEDA-AC) used in the air cleaning unit of nuclear power plants (NPPs) should be maintained at least 99% between 24 month-performance test period. In order for evaluating the effectiveness of TEDA-AC on the removal performance of CH₃I in nuclear power plant during the operation of NPPs, the long-term test for up to 15 months was carried out under the simulated operating conditions (e.g., 25 ℃, RH 50%, ppb level poisoning gases injection) at nuclear power plants (NPPs). The TEDA-AC samples were analyzed with the Brunauer-Emmett-Teller (BET) specific surface area and TEDA content as well as CH₃I penetration test. It is clearly evident that more than 99% of CH₃I removal performance of TEDA-AC was observed in the TEDA-AC samples during 15 months of long-term operation under the simulated NPP operating conditions including the ppb level of organic and oxide form of poisoning gases. BET specific surface area and TEDA content that can affect the CH₃I removal performance of TEDA-AC were also maintained as those in new TEDA-AC during 15 months of long-term operation.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.153-163
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• 2016

Environments in nuclear power plants (NPPs) are changing as the design of instrumentation and control systems for NPPs is rapidly moving toward fully digital instrumentation and control, and modern computer techniques are gradually introduced into main control rooms (MCRs). Within the context of these environmental changes, the level of performance of operators in a digital MCR is a major concern. Situation awareness (SA), which is used within human factors research to explain to what extent operators of safety-critical systems know what is transpiring in the system and the environment, is considered a prerequisite factor to guarantee the safe operation of NPPs. However, the safe operation of NPPs can be guaranteed through a team effort. In this regard, the operating team's SA in a conventional and digital MCR should be measured in order to assess whether the new design features implemented in a digital MCR affect this parameter. This paper explains the team SA measurement method used in this study and the results of applying this measurement method to operating teams in different MCR environments. The paper also discusses several empirical lessons learned from the results.

• Journal of the Ergonomics Society of Korea
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• v.31 no.1
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• pp.261-269
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• 2012

Objective: The aim of this study is to develop an analysis method to extract plausible types of errors when using a smart mobile in nuclear power plants. Background: Smart mobiles such as a smart-phone and a tablet computer(smart-pad) are to be introduced to the various industries. Nuclear power plant like APR1400 already adopted many up-to-date digital devices within its main control room. With this trend, various types of smart mobiles will be inevitably introduced to the nuclear field in the near future. However nuclear power plants(NPPs) should be managed considering a big risk as a result of the trend not only economically but also socially compared to the other industrial systems. It is formally required to make sure to reasonably prevent the all hazards due to the introduction of new technologies and devices before the application to the specific tasks in nuclear power plants. Method: We define interaction segments(IS) as a main architect of interaction description, and enumerate all plausible error segments(ES) for a part of design evaluation of digital devices. Results: We identify various types of interaction errors which are coped with reasonably by interaction design using smart mobiles. Conclusion: According to the application result of the proposed method, we conclude that the proposed method can be utilized to specify the requirements to the human error hazards in digital devices, and to conduct a human factors review during the design of digital devices. Application: The proposed method can be applied to predict the human errors of the tasks related to the digital devices; therefore we can ensure the safety to apply the digital devices to be introduced to NPPs.