• Nuclear Engineering and Technology
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• v.43 no.1
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• pp.45-62
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• 2011

The Korean nuclear industry is developing a thermal-hydraulic analysis code for safety analysis of pressurized water reactors (PWRs). The new code is called the Safety and Performance Analysis Code for Nuclear Power Plants (SPACE). The SPACE code adopts advanced physical modeling of two-phase flows, mainly two-fluid three-field models which comprise gas, continuous liquid, and droplet fields and has the capability to simulate 3D effects by the use of structured and/or nonstructured meshes. The programming language for the SPACE code is C++ for object-oriented code architecture. The SPACE code will replace outdated vendor supplied codes and will be used for the safety analysis of operating PWRs and the design of advanced reactors. This paper describes the overall features of the SPACE code and shows the code assessment results for several conceptual and separate effect test problems.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4743-4750
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• 2022

Heat pipe cooled nuclear reactor has been a very attractive technical solution to provide the power for deep space applications. In this paper, a 200 kWe space nuclear reactor power design has been proposed based on the combination of an integrated UN ceramic fuel, a heat pipe cooling system and the Stirling power generators. Neutronics and thermal analysis have been performed on the space nuclear reactor. It was found that the entire reactor core has at least 3.9 $ subcritical even under the worst-case submersion accident superimposed a single safety drum failure, and results from fuel temperature coefficient, neutron spectrum and power distribution analysis also showed that this reactor design satisfies the neutronics requirements. Thermal analysis showed that the power in the core can be successfully removed both in normal operation or under one or more heat pipes failure scenarios.

• Journal of the Korean Society of Safety
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• v.19 no.2
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• pp.11-15
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• 2004

The fretting-fatigue by leaking is one of the significant degradation in steam generator tubes. In this study, the burst pressure of inclined damaged steam generator tubes were obtained from three criterions by using the finite element method. The analysis results were also compared with the experiment data from published references and they showed a good agreement with the experiment data.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.494-495
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• 2005

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.997-1005
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• 2018

The regulatory periodic safety inspection system is one of the most important methods for confirming the safety of nuclear power plants and the defense in depth in nuclear safety is the most important basic means for accident prevention and mitigation. Recently, a new regulatory technology based on risk-informed and safety performance has been developed and used in advanced countries. However, since the domestic periodic inspection system is being used in the same way over 30 years, it is necessary to know how the inspection contributes to the safety confirmation of the nuclear power plants. In this study, the domestic periodic inspection system currently in use was analyzed from the perspective of defense in depth in nuclear safety. In addition, the analysis results were compared to the U.S. NRC's safety inspection system to obtain consistency and lessons in this study. As a result of analysis, the NRC's safety inspections were distributed almost evenly at the all levels of defense in depth, while in the case of domestic inspection, they were heavily focused on the level 1 of defense in depth. Therefore, it appeared urgent to improve the inspection system to strengthen the other levels of defense in depth in nuclear safety.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.1117-1131
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• 2019

Tanks are used extensively in many engineering areas for spent fuel pool structures at nuclear power plants or for water storage tanks in bulk carriers. To ensure the structural integrity of such tanks when under dynamic loads, modal characteristics such as natural frequencies, participation factors and mode shapes should be known. Investigated in this study are the modal characteristics of a square tank by the finite element method. This approach can be used with subsequent dynamic analyses such as a response spectrum analysis or a harmonic analysis. Finite element models are prepared to determine the natural frequencies and mode shapes, which are easy to find the modal characteristics of a fluid-filled square tank. The effects of the fluid contained in the tank and the boundary conditions at top and bottom ends on the modal characteristics are assessed by several finite element analyses.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1696-1707
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• 2019

This paper conducts a qualitative policy analysis of current challenges to safety culture and security culture in Southeast Asia and emerging best practices in Northeast Asia that are aimed at strengthening both cultures. It analyses lessons, including strengths and limitations, that can be derived from Northeast Asian states, given the long history of nuclear energy in South Korea, China and Japan. It identifies and examines best practices from Northeast Asia's Nuclear Security Centres of Excellence in terms of boosting nuclear security culture and their relevance for Southeast Asia. The paper accentuates the important role of the State in adopting policy and regulatory frameworks and in institutionalising nuclear education and training programmes to deepen the safety-security cultures. Best practices in and challenges to developing a nuclear safety culture and a security culture in East Asia are examined using three frameworks of analysis (i) a comprehensive nuclear policy framework; (ii) a proactive and independent regulatory body; and (iii) holistic nuclear education and training programmes. The paper argues that Southeast Asian states interested in harnessing nuclear energy and/or utilising radioactive sources for non-power applications must develop a comprehensive policy framework on developing safety and security cultures, a proactive regulatory body, and holistic nuclear training programmes that cover both technical and human factors. Such measures are crucial in order to mitigate human errors that may lead to radiological accidents and nuclear security crises. Key lessons from Japan, South Korea and China such as best practices and challenges can inform policy recommendations for Southeast Asia in enhancing safety-security cultures.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.676-682
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• 2019

After the Fukushima Daiichi accident, there has been an increasing preference for passive safety features in the nuclear power industry. Some passive safety systems require limited active components to trigger subsequent passive operation. Under very serious accident conditions, passive safety features could be rendered inoperable or damaged. This study evaluates (i) the performance and effectiveness of the passive safety features of iPOWER (innovative Power Reactor), and (ii) whether a severe accident condition could be reached if the passive safety systems are damaged, namely the case of heat exchanger tube rupture. Analysis results show that the reactor coolant system remains in the hot shutdown condition without operator actions or electricity for over 72 h when the passive auxiliary feedwater systems (PAFSs) are operable without damage. However, heat exchanger tube rupture in the PAFS leads to core damage after about 18 h. Such results demonstrate that, to enhance the safety of iPOWER, maintaining the integrity of the PAFS is critical, and therefore additional protections for PAFS are necessary. To improve the reliability of iPOWER, additional battery sets are necessary for the passive safety systems using limited active components for accident mitigation under such extreme circumstances.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2067-2076
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• 2022

The importance of safety culture has been emphasized to achieve a high level of safety. In this light, a systematic method to more properly deal with safety culture is necessary. Here, a decision-making tool that can apply a graded approach to the analysis of safety culture is proposed, called the F-D matrix, which determines the frequency and the difficulty of safety culture attributes recently defined by the IAEA. A hierarchical model of difficulty contributors was developed as a scoring standard, and its elements were weighted via expert evaluation using the analytic hierarchy process. The frequency of the attributes was derived by analyzing reported events from nuclear power plants in the Republic of Korea. Period-by-period comparisons with the F-D matrix can show trends in the change of the maturity level of an organization's safety culture and help to evaluate the effectiveness of previously implemented measures. In the evaluating the difficulty of the attributes in the recently developed harmonized safety culture model, the difficulties of Trending, Benchmarking, Resilience, and Documentation and Procedures were found to be relatively high, while the difficulties of Conflicts are Resolved, Ownership, Collaboration, and Respect is Evident were found to be relatively low. A case study was conducted with an analysis period of 10 years to attempt to reflect the many changes in safety culture that have been made following the Fukushima accident in March 2011. As a result of comparing two periods following the Fukushima accident, the overall frequency decreased by about 40%, providing evidence for the effects of the various improvements and measures taken following the increased emphasis on safety culture. The proposed F-D matrix provides a new analytical perspective and enables an in-depth analysis of safety culture.

• Nuclear Engineering and Technology
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• v.41 no.1
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• pp.91-98
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• 2009

The "3+3 Process" for safety critical software for nuclear power plants' I&C (Instrumentation and Control system) has been developed in this work. The main idea of the "3+3 Process" is both to simplify the software development and safety analysis in three steps to fulfill the requirements of a software safety plan [1]. The "3-Step" software development process consists of formal modeling and simulation, automated code generation and coverage analysis between the model and the generated source codes. The "3-Step" safety analysis consists of HAZOP (hazard and operability analysis), FTA (fault tree analysis), and DV (design validation). Put together, these steps are called the "3+3 Process". This scheme of development and safety analysis minimizes the V&V work while increasing the safety and reliability of the software product. For assessment of this process, validation has been done through prototyping of the SDS (safety shut-down system) #1 for PHWR (Pressurized Heavy Water Reactor).