• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.1959-1974
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• 2024

Nuclear energy plays a crucial role in energy supply in the 21st century, and more and more Nuclear Power Plants (NPPs) will be in operation to contribute to the development of human society. However, as a typical complex system engineering, the operation and development of NPPs require efficient and stable control methods to ensure the safety and efficiency of nuclear power generation. Reinforcement learning (RL) aims at learning optimal control policies via maximizing discounted long-term rewards. The reward-oriented learning paradigm has witnessed remarkable success in many complex systems, such as wind power systems, electric power systems, coal fire power plants, robotics, etc. In this work, we try to present a systematic review of the applications of RL on these complex systems, from which we believe NPPs can borrow experience and insights. We then conduct a block-by-block investigation on the application scenarios of specific tasks in NPPs and carried out algorithmic research for different situations such as power startup, collaborative control, and emergency handling. Moreover, we discuss the possibilities of further application of RL methods on NPPs and detail the challenges when applying RL methods on NPPs. We hope this work can boost the realization of intelligent NPPs, and contribute to more and more research on how to better integrate RL algorithms into NPPs.

• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.820-830
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• 2016

Pipe wall thinning by flow-accelerated corrosion and various types of erosion is a significant and costly damage phenomenon in secondary piping systems of nuclear power plants (NPPs). Most NPPs have management programs to ensure pipe integrity due to wall thinning that includes periodic measurements for pipe wall thicknesses using nondestructive evaluation techniques. Numerous measurements using ultrasonic tests (UTs; one of the nondestructive evaluation technologies) have been performed during scheduled outages in NPPs. Using the thickness measurement data, wall thinning rates of each component are determined conservatively according to several evaluation methods developed by the United States Electric Power Research Institute. However, little is known about the conservativeness or reliability of the evaluation methods because of a lack of understanding of the measurement error. In this study, quantitative models for UT thickness measurement deviations of nuclear pipes and fittings were developed as the first step for establishing an optimized thinning evaluation procedure considering measurement error. In order to understand the characteristics of UT thickness measurement errors of nuclear pipes and fittings, round robin test results, which were obtained by previous researchers under laboratory conditions, were analyzed. Then, based on a large dataset of actual plant data from four NPPs, a quantitative model for UT thickness measurement deviation is proposed for plant conditions.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2888-2898
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• 2021

Vital area identification (VAI) is an essential procedure for the design of physical protection systems (PPSs) for nuclear power plants (NPPs). The purpose of PPS design is to protect vital areas. VAI has been improved continuously to overcome the shortcomings of previous VAI generations. In first-generation VAI, a sabotage fault tree was developed directly without reusing probabilistic safety assessment (PSA) results or information. In second-generation VAI, VAI model was constructed from all PSA event trees and fault trees. While in third-generation VAI, it was developed from the simplified PSA event trees and fault trees. While VAIs have been performed for NPPs in full-power operations, VAI for NPPs in low-power and shutdown (LPSD) operations has not been studied and performed, even though NPPs in LPSD operations are very vulnerable to sabotage due to the very crowded nature of NPP maintenance. This study is the first to research and apply VAI to LPSD operation of NPP. Here, the third-generation VAI method for full-power operation of NPP was adapted to the VAI of LPSD operation. In this study, LPSD VAI for a few plant operational states (POSs) was performed. Furthermore, the operation strategy of vital areas for both full-power and LPSD operations was discussed. The LPSD VAI method discussed in this paper can be easily applied to all POSs. The method and insights in this study can be important for future LPSD VAI that reflects various LPSD operational states. Regulatory bodies and electric utilities can take advantage of this LPSD VAI method.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.70-74
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• 2014

An Operating Experience Report(OER) has written about events and accidents happened at a Nuclear Power Plant(NPP). The purpose of publishing the OER is to prevent the similar event or accident repeatedly by spreading the experience of a single plant to other plants personnel. In this paper, it is analyses that the foreign NPPs' OERs on JIT published by the International Nuclear Agency(WANO, INPO, COG, BE). The analysis introduced in this paper is performed along with the various factors such as type of work, root-cause, and equipment. The root-cause analysis about the OERs shows that the Human-error is the major factor in foreign NPPs, but on the other hand equipment problem is the main part of the Domestic NPPs. The ratio of the foreign NPP's OERs on JIT according to the type of work was applied to KHNP-JIT developed nowadays for the first time in KOREA.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.351-356
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• 2021

Korean nuclear power plants (NPPs) have various radiation protection programs to attain radiation exposure as low as reasonably achievable (ALARA). In terms of ALARA, this paper provides a comprehensive overview of administrative dose control for occupationally-exposed workers in Korean NPPs. In addition to dose limits, administrative dose constraints are implemented to resolve an inequity of radiation exposure in which some individuals in NPPs receive relatively higher doses than others. Occupational dose constraints in Korean NPPs are presented in this paper with the background of how those values were determined. For pressurized water reactors, 80% and 90% of the annual average limit for an effective dose, 20 mSv/y, are set as the primary and secondary dose constraints, respectively. Pressurized heavy water reactors (PHWRs) have also established the primary and secondary dose constraints corresponding to 70% and 80% of the effective dose limit, and additional constraints for tritium concentration are provided to control internal exposure in PHWRs. Follow-up measures for exceeding these administrative dose constraints are also introduced compared to exceeding the dose limits. Finally, analysis results of dose distributions show how the implementation of administrative dose constraints impacted the occupational dose distributions in Korean NPPs during the years 2009-2018.

• International conference on construction engineering and project management
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• 2020.12a
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• pp.353-358
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• 2020

Recent efforts to develop a common standard for nuclear power plants (NPPs) with the aim of creating (1) a digital environment for a better understanding of NPPs life-cycle management aspect and (2) engineering data interoperability by using existing standards among different unspecified project participants (e.g., owners/operators, engineers, contractors, equipment suppliers) during plants' life cycle process (EPC, O&M, and decommissioning). In order to meet this goal, there is a need for formulating a standardized high-level physical breakdown structure (PBS) for NPPs project management office (PMO). However, high-level PBS must be comprehensive enough and able to represent the different types of plants and the new trends of technologies in the industry. This has triggered the need for addressing the issues of the recent operational NPPs and future technologies' ramification for evaluating the changes in the NPPs physical components in terms of structure, system, and component (SSC) configuration. In this context, this ongoing study examines the recent conventional NPPs and technological trends in the development of future NPPs facilities. New reactor models regarding the overlap of variant issues of nuclear technology were explored. Finally, issues on PBS for project management are explored by the examination of the configuration of NPPs primary system. The primary systems' configuration of different reactor models is assessed in order to clarify the need for analyzing the new trends in nuclear technology and to formulate a common high-level PBS. Findings and implications are discussed for further studies.

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

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.35-39
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• 2013

The Operating Experience Report(OER) has written about the event and accident happened at a Nuclear Power Plant(NPP). The purpose of publishing the OER is to prevent the similar event or accident repeatedly by spreading the experience of a single plant to other plants personnel. Before initiating the analysis mentioned in this paper, 2,298 review reports for the same number of OER published from 2007 to June 2012 have been written to achieve the correct and objective statistics. The analysis introduced in this paper is performed with the various factors such as year, plant type, equipment, type of work, root-cause. The root-cause analysis is showed that the equipment problem is the major factor in domestic NPPs, but on the other hand human-error is the main part of the foreign NPPs. Moreover, while the number of the man-made event is decreasing, the equipment-made event is rapidly increasing in domestic NPPs.

• Nuclear Engineering and Technology
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• v.49 no.8
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• pp.1772-1777
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• 2017

Korean nuclear power plants (NPPs) periodically evaluate the radioactive gaseous and liquid effluents released from power reactors to protect the public from radiation exposure. This paper provides a comprehensive overview of the release of radioactive effluents from Korean NPPs and the effects on the annual radiation doses to the public. The amounts of radioactive effluents released to the environment and the resulting radiation doses to members of the public living around NPPs were analyzed for the years 2011-2015 using the Korea Hydro & Nuclear Power Co., Ltd's annual summary reports of the assessment of radiological impact on the environment. The results show that tritium was the primary contributor to the activity in both gaseous and liquid effluents. The averages of effective doses to the public were approximately on the order of $10^{-3}mSv$ or $10^{-2}mSv$. Therefore, even though Korean NPPs discharged some radioactive materials into the environment, all effluents were within the regulatory safety limits and the resulting doses were much less than the dose limits.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.88-96
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• 2018

Concerns over reliability assessments of the main components in nuclear power plants (NPPs) related to aging and continuous operation have increased. The conventional reliability assessment for main components uses experimental correlations under general conditions. Most NPPs have been operating in Korea for a long time, and it is predictable that NPPs operating for the same number of years would show varying extent of aging and degradation. The conventional reliability assessment does not adequately reflect the characteristics of an individual plant. Therefore, the reliability of individual components and an individual plant was estimated according to operating data and conditions. It is essential to reflect aging as a characteristic of individual NPPs, and this is performed through prognostics. To handle this difficulty, in this paper, the general path model/Bayes, a data-based prognostic method, was used to update the reliability estimated from the generic database. As a case study, the authors consider the aging for steam generator tubes in NPPs and demonstrate the suggested methodology with data obtained from the probabilistic algorithm for the steam generator tube assessment program.