• Nuclear Engineering and Technology
• /
• v.55 no.11
• /
• pp.4252-4265
• /
• 2023

In the case of nuclear power plants near end of their design life, a reassessment of the performance of safetyrelated equipment may be necessary to determine whether to shut down or extend the operation of the power plant. Therefore, it is necessary to evaluate the level of performance decline due to degradation. Electrical cabinets, including MCC and switchgear, are representative safety-related equipment. Several studies have assessed the degradation and seismic performance of nuclear power plant equipment. Most of those researches are limited to individual components due to the size of safety-related equipment and test equipment. However, only a few studies assessed the degradation performance of electrical cabinets. The equipment of various nuclear power plants is anchored to concrete foundations, and crack in concrete foundations is one of the most representative of degradation that could be visually confirmed. However, it is difficult to find a study for analysis through testing the effect of cracks in concrete foundations on the response of electrical cabinet internal equipment fixed by anchors. In this study, using a simple cabinet model considering cast-in-place anchor in uncracked and cracked concretes, a tri-axial shaking table tests were performed and the seismic behavior were observed.

• Nuclear Engineering and Technology
• /
• v.56 no.3
• /
• pp.1116-1123
• /
• 2024

-In recent years, unmanned underwater vehicles (UUV) have been vigorously developed, and with the continuous deepening of marine exploration, traditional energy can no longer meet the energy supply. Nuclear energy can achieve a huge and sustainable energy supply. The heat pipe reactor has no flow system and related auxiliary systems, and the supporting mechanical moving parts are greatly reduced, the noise is relatively small, and the system is simpler and more reliable. It is more favorable for the control of unmanned systems. The use of heat pipe reactors in unmanned underwater vehicles can meet the needs for highly compact, long-life, unmanned, highly reliable, ultra-quiet power supplies. In this paper, a heat pipe reactor scheme named UPR-S that can be applied to unmanned underwater vehicles is designed. The reactor core can provide 1 MW of thermal power, and it can operate at full power for 5 years. UPR-S has negative reactive feedback, it has inherent safety. The temperature and stress of the reactor are within the limits of the material, and the core safety can still be guaranteed when the two heat pipes are failed.

• Nuclear Engineering and Technology
• /
• v.53 no.11
• /
• pp.3741-3758
• /
• 2021

Outcomes of the FALCON code analysis-related part of the STARS-ENSI Service Project on Evaluation of the new U.S.NRC RIA Fuel Safety Criteria and Application to the Swiss Reactors are presented. Substantial conservatism of the updated safety limits for high-temperature and PCMI cladding failure, as proposed in the NRC Regulatory Guide RG 1.236, is confirmed. Applicability of the updated failure limits to fuel safety analysis in the Swiss PWRs, as applied to standard fuel designs using UO2 fuel pellets and SRA Zry-4 as cladding materials is discussed. Conducting of new integral RIA tests with irradiated samples using doped- and gadolinia fuel pellets to support appropriate fuel safety criteria for RIA events is recommended.

• Nuclear Engineering and Technology
• /
• v.53 no.10
• /
• pp.3319-3326
• /
• 2021

As a form of industrial control systems (ICS), nuclear instrumentation and control (I&C) systems have been digitalized increasingly. This has raised in turn cyber security concerns. Cyber security for ICS is important because cyber-attacks against ICS can cause not only equipment damage and loss of production but also personal and public safety hazards unlike in general IT environments. Numerous risk analyses have been carried out to enhance the safety of ICS and recently, many studies related to the cyber security of ICS are being conducted. Many existing risk analyses and cyber security studies have considered safety and cyber security separately. However, both safety and cyber security perspectives should be considered when analyzing risks for complex and critical ICS facilities such as nuclear power plants (NPPs). In this paper, the STPA-SafeSec methodology is selected to consider both safety and security perspectives when performing a risk analysis for NPPs in order to assess impacts on the safety by cyber-attacks against the digital I&C systems. The STPA-SafeSec methodology was applied to a test-bed system that simulates a condensate water (CD) system in an NPP. The process of the application up to the development of mitigation strategies is described in detail.

• Nuclear Engineering and Technology
• /
• v.45 no.2
• /
• pp.125-140
• /
• 2013

This paper aims to give an overview of the methods to inherently prevent human errors and to effectively mitigate the consequences of such errors by securing defense-in-depth during plant management through the advanced man-machine interface system (MMIS). It is needless to stress the significance of human error reduction during an accident in nuclear power plants (NPPs). Unexpected shutdowns caused by human errors not only threaten nuclear safety but also make public acceptance of nuclear power extremely lower. We have to recognize there must be the possibility of human errors occurring since humans are not essentially perfect particularly under stressful conditions. However, we have the opportunity to improve such a situation through advanced information and communication technologies on the basis of lessons learned from our experiences. As important lessons, authors explained key issues associated with automation, man-machine interface, operator support systems, and procedures. Upon this investigation, we outlined the concept and technical factors to develop advanced automation, operation and maintenance support systems, and computer-based procedures using wired/wireless technology. It should be noted that the ultimate responsibility of nuclear safety obviously belongs to humans not to machines. Therefore, safety culture including education and training, which is a kind of organizational factor, should be emphasized as well. In regard to safety culture for human error reduction, several issues that we are facing these days were described. We expect the ideas of the advanced MMIS proposed in this paper to lead in the future direction of related researches and finally supplement the safety of NPPs.

• Journal of Energy Engineering
• /
• v.23 no.2
• /
• pp.191-198
• /
• 2014

Korea Hydro & Nuclear Power Co. (KHNP) has developed and implemented the maintenance effectiveness monitoring (MR) programs for the operating nuclear power plants. MR programs are developed by reflecting design characteristics of the operating nuclear power plants to monitor the plant performance for improving the safety and reliability. Recently, KHNP has developed the MR program for APR1400 safety related systems to establish the advanced maintenance system and will verify the suitability of the MR program through evaluating initial performance. Consequently, it is expected that the safety of the new plant will be improved by developing and implementing the MR program.

• Corrosion Science and Technology
• /
• v.14 no.1
• /
• pp.25-32
• /
• 2015

The alloy content of structural materials of nuclear power plants has been recognized an important factor in predicting flow accelerated corrosion (FAC). In particular, many literature data reported that chromium content is one of the most important alloying element and even a small amount of chromium is effective to suppress FAC. This report reviewed and compared chromium models of Ducreux, Bouchacourt, and Kastner which were used in predicting FAC rates. The plant data indicate that Ducreux model may be conservative for the specimen containing 0.15 wt% chromium. The related articles were reviewed as follows. Combined effects of chromium content, pH, temperature, dissolved oxygen (DO), flow velocity, test time, and kinds of amine on the FAC rate were described. 0.1 wt% chromium in steel did not affect the FAC rate with changes in pH. The FAC rates pronounced with higher flow rate and increased with increasing test duration(600 d) for 0.013 wt% chromium. The FAC rates in mixed amine chemistry were higher than in ammonia chemistry, which may be lessened by the addition of chromium to the steel.

• Nuclear Engineering and Technology
• /
• v.41 no.4
• /
• pp.545-548
• /
• 2009

Korean utility has utilized a Periodic Safety Review (PSR) that assesses the cumulative effects of plant aging, modifications, operating experience, technical developments, and site characteristics since 2000. In particular, the assessment and management of plant aging is one of the major areas in PSR. It includes identification of critical Systems, Structures, and Components (SSCs) for aging, assessment of aging effects, and implementation of aging management programs. Since the PSR system was introduced based on the atomic energy acts and related laws, PSRs of eight sets for 12 Nuclear Power Plants (NPPs) that have been operating more than 10 years have been completed. PSRs of two sets for 4 NPPs are currently being carried out. The utility has confirmed that domestic NPPs have been operated safely through these PSRs and have implemented the follow-up corrective activities to increase the nuclear safety. In this paper, the status of PSR implementation is discussed and improvement programs to conduct PSR follow-up corrective activities efficiently for NPPs are suggested based on experiences with aging assessments.

• Journal of Digital Convergence
• /
• v.17 no.8
• /
• pp.47-60
• /
• 2019

The purpose of this study is to explore the overall structural relationship between the statutory provisions of nuclear energy legislation and to identify the coherence of the nuclear law system using social network analysis. In particular, we analyze the legal structure of the "Nuclear Safety Act", which plays a central role in nuclear safety regulation, to examine the key provisions in legal network structure of Nuclear Safety Act. Therefore, we found the structural problems of the nuclear legal system and suggest the legislative improvement plan for reducing excessive legislative activity and determining the need for legal amendments in nuclear safety management and regulation. This study is expected to provide a analytical framework for making legal system of further policy in other science and technology industries as well as nuclear energy related industries.

• Nuclear Engineering and Technology
• /
• v.45 no.2
• /
• pp.165-178
• /
• 2013

A comprehensive review on the technical standards about human factors (HF) design and software reliability maintenance for digital instrumentation and control (I&C) and human-machine interface technology (HMIT) in Japanese light water reactor nuclear power plants (NPPs) was given in this paper mainly by introducing the relevant activities at the Japan Electric Association to set up many industrial standards within the traditional framework of nuclear safety regulation in Japan. In Japan, the Fukushima Daiichi accident that occurred on March 11, 2011 has great impact on nuclear regulation and nuclear industries where concerns by the general public about safety have heightened significantly. However for the part of HF design and software reliability maintenance of digital I&C and HMIT for NPP, the author believes that the past practice of Japanese activities with the related technical standards can be successfully inherited in the future, by reinforcing the technical preparedness for the prevention and mitigation against any types of severe accident occurrence.