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Aoya, Juri; Miyata, Katsuhiko*; Terakado, Akihito*; Otsuzumi, Yoji*; Kurosawa, Daiki*; Sunaba, Takanobu*; Oyama, Yuto*; Inada, Satoshi
Nihon Hozen Gakkai Dai-17-Kai Gakujutsu Koenkai Yoshishu, p.507 - 512, 2021/07
The high level radioactive liquid waste is analyzed for the vitrification process control and the vitrified waste quality in the hot cell of Tokai Vitrification Facility, Tokai Reprocessing Plant. There are 8 Master-slave manipulators, 7 lighting equipment, an electronic balance, and an inductively coupled plasma-optical emission spectrometer used for remote operation, securing visibility, total oxide analysis, and elemental analysis in the analytical hot cell. These equipment and analytical apparatus must be secured with the integrity all the time because the vitrification process cannot be proceeded without analysis of the high level radioactive liquid waste. We constructed the self-remote-maintenance technologies of these equipment and analytical apparatus which reduce the risks of radioactive contamination, radiation exposure, and injury of an operator and also were optimized with respect to a labor, time, and cost, based on the operation of approximately 20 years.
Taguchi, Shigeo; Taguchi, Katsuya; Makino, Risa; Yamanaka, Atsushi; Suzuki, Kazuyuki; Takano, Masato; Koshino, Katsuhiko; Ishida, Michihiko; Nakano, Takafumi; Yamaguchi, Toshiya
Nihon Hozen Gakkai Dai-17-Kai Gakujutsu Koenkai Yoshishu, p.499 - 502, 2021/07
In 2018, Tokai Reprocessing Plan (TRP) shifted to the decommissioning stage. In order to proceed with steady decommissioning work, TRP effort to enhance project management function. This paper describes the establishment and role of the Decommissioning Project Management Office, effectiveness of applying the project management tool and its utilization concept, and the method of materialize the equipment dismantling plan.
Nishino, Saki; Tsuboi, Masatoshi; Okada, Jumpei; Saegusa, Yu; Omori, Kazuki; Yasuo, Kiyoshi; Seshimo, Kazuyoshi; Domura, Kazuyuki; Yamamoto, Masahiko
Nihon Hozen Gakkai Dai-17-Kai Gakujutsu Koenkai Yoshishu, p.541 - 548, 2021/07
no abstracts in English
Goto, Sho; Aoki, Kenji; Morimoto, Kenji; Tsuboi, Masatoshi; Isozaki, Naohiko; Furukawa, Ryuichi; Kitagawa, Osamu; Fukaya, Yasuhiro*
Nihon Hozen Gakkai Dai-17-Kai Gakujutsu Koenkai Yoshishu, p.517 - 520, 2021/07
no abstracts in English
Funakoshi, Tomomasa; Shibata, Atsuhiro; Kitawaki, Shinichi; Yajima, Hiroshi*; Masaoka, Hideki*; Koka, Masashi*
no journal, ,
no abstracts in English
Takase, Yuki; Miura, Yasushi; Nakabayashi, Hiroki
no journal, ,
Tokai Reprocessing plant (TRP), which temporarily suspended the reprocessing of spent fuel in 2007, moved to decommissioning in 2018 without restarting operations. Under the decommissioning situation, the high-level radioactive liquid waste generated in the past reprocessing operation is the highest potential risk in TRP. These high-level radioactive liquid wastes are stored in High Active Liquid Waste Storage Facility (HAW) and sequentially Vitrified in Tokai Vitrification Facility (TVF). The serious accident risk about evaporation accident of high radioactive liquid waste has been remained in both facilities. In this study, a loss of heat sink (LOHS) frequency of high radioactive liquid waste in HAW and TVF was assessed by Probability Risk Assessment (PRA). The risk information was extracted for the equipment of important of safety related system from the view point of maintaining the facility safety. Its capability to maintenance planning was evaluated.
Aizawa, Kosuke; Ishimaru, Masaru; Maeda, Shigetaka; Okada, Taiichi*; Matsui, Hiroyuki*; Kumagai, Yukihiro*; Isaka, Shingo*
no journal, ,
The protective nets against the tornado missile are important for power plant equipment. Japan Atomic Energy Agency has developed the protective net using aramid fiber rope to aim high strength lightweight net. The net element tests and analyses were conducted to clarify the characteristic of the protective net using aramid fiber rope. The actual size net was designed based on the results of net element tests and analyses. The dynamic analysis which simulates a car crash on the designed actual size net was conducted. From the result of the dynamic analysis, it was confirmed that the protective net using aramid fiber rope has a possibility to withstand the car crash. In addition, the fire resistance test and the resistance to weather of the aramid fiber rope were conducted. The results showed that the aramid fiber rope has high performance for firing resistance and weather resistance.
Onizawa, Toshikazu; Hoshino, Masato; Nemoto, Hidenori; Akiyama, Kazuki
no journal, ,
Low level liquid waste generated from reprocessing of spent nuclear fuel in the Tokai Reprocessing Plant solidified with bitumen from 1982 to 1997. About 30,000 of bituminized wastes were produced, and stored in the storage facility. Recently, the leakage phenomenon due to the corrosion of bituminized waste drum occurred in the other nuclear facility. Based on the malfunction event, we have advanced the development of observation device for grasping soundness of the bituminized waste drums stored in the storage facility. This paper is reported about our effort to develop a remote observation device of the bituminized waste drums stored in the storage facility and the observation results with the developed device.
Toyota, Kodai; Hashidate, Ryuta; Yada, Hiroki; Takaya, Shigeru; Miyakoshi, Hiroyuki; Kato, Shoichi
no journal, ,
Probabilistic fracture mechanics (PFM) evaluation requires information on the probabilistic distributions of the number and size of initial defects, material properties such as crack growth rate due to fatigue and creep, and load to evaluate the failure probability of components. In this study, ultrasonic testing was conducted on the welds of the test equipment used in the research and development of fast reactors, and the number and size of defects were evaluated. The results could be used as conservative values of initial defects, and the values related to initial defects for PFM evaluation of FBR components were examined.
Hashidate, Ryuta; Yada, Hiroki; Takaya, Shigeru; Futagami, Satoshi; Yamano, Hidemasa; Kurisaka, Kenichi; Enuma, Yasuhiro
no journal, ,
If components can be maintained according to a target reliability of nuclear power plants, plant safety and economic efficiency could be improved by avoiding excessive or insufficient margins. This approach can rationalize the maintenance of low target reliability components and concentrate resources (people, money, and time) on the maintenance of high target reliability components. It would take a long time to evaluate the target reliability for all components in the plant by quantitatively using probabilistic risk assessment. This paper, therefore, proposed at a first step a simple screening method which determines necessity of quantitative risk assessment for maintenance rationalization using risk information. Specifically, this method was applied to determining necessity of quantitative risk assessment to the laundry system whose function has little contributor to the plant safety of nuclear power plant, and we proposed the maintenance rationalization of the laundry system.
Kondo, Yuki; Hashidate, Ryuta; Hamase, Erina; Ezure, Toshiki; Mitsumoto, Rika; Yada, Hiroki; Enuma, Yasuhiro
no journal, ,
JAEA has started the development of Advanced Reactor Knowledge-and AI-aided Design Integration Approach through the whole plant lifecycle (ARKADIA) that performs integrated design evaluation and optimization from various viewpoints such as safety, economy, and maintainability using risk information, in order to dramatically improve the development efficiency of innovative reactors. In this paper, we report the concept of a knowledge management system to support integrated design evaluation and optimization, and the realization of the knowledge base necessary for the design and optimization of innovative reactors.
Suzuki, Masaaki*; Ito, Mari*; Hashidate, Ryuta; Takahashi, Keita; Yada, Hiroki; Takaya, Shigeru
no journal, ,
Maintenance scheduling is currently manually handled, which is a time-consuming process because of the large number of components and constraints that must be taken into account when creating a schedule. Besides, to develop next-generation power plants with excellent operability, it is necessary to make it possible to evaluate operability and maintainability in advance at the design stage. Our objective is to develop and implement an automatic scheduling system using the mathematical technique of Operations Research for addressing the inspection process scheduling problem in a sodium-cooled fast reactor plant. This study constructs a scheduling model that performs optimization in two stages to reduce the computation costs.
Dong, F.*; Chen, S.*; Demachi, Kazuyuki*; Hashidate, Ryuta; Takaya, Shigeru
no journal, ,
Piping and Instrumentation Diagrams contain information about the piping and process equipment together with the instrumentation and control devices, which is essential to the design and management of Nuclear Power Plants. There are abundant complex objects on P&IDs, with imbalanced distribution of these objects and their linked information across different diagrams. Therefore, the content of P&IDs is generally extracted and analyzed manually, which is time consuming and error prone. To efficiently address these issues, we integrate state-of-the-art deep learning-based object detection and Optical Character Recognition models to automatically extract link information from P&IDs. Besides, we propose a novel image pre-processing approach using sliding windows to detect low resolution small objects. The performance of the proposed approach was experimentally evaluated, and the experimental results demonstrate it capable to extract link information from P&IDs of NPPs.
Kuwabara, Yuto*; Demachi, Kazuyuki*; Kasahara, Naoto*; Chen, S.*; Nishino, Hiroyuki; Onoda, Yuichi; Kurisaka, Kenichi
no journal, ,
In order to quantitatively evaluate the ability of a nuclear plant to recover its safety functions, we are developing a method to simulate accident management in chronological order according to an accident scenario, rather than simply evaluating the probability, and to evaluate whether or not a major accident will eventually occur, i.e., whether or not the minimum necessary safety functions can be recovered within a time limit. In this presentation, we will discuss the development of a method to evaluate whether or not the minimum necessary safety functions can be recovered within the time limit. In this presentation, the specific procedure and management examples of the method will be explained.
Demachi, Kazuyuki*; Kuwabara, Yuto*; Kasahara, Naoto*; Nishino, Hiroyuki; Onoda, Yuichi; Kurisaka, Kenichi
no journal, ,
Our aim is to develop a technology to suppress the expansion of accident damage by improving the reactor structural resilience as a solution to the problem of restoring the safety function of structures after destruction, which has been an issue since the Fukushima Daiichi Nuclear Power Plant accident. In this research, the visualization method of resilience of nuclear structures was proposed in order to visualize the capacity to mitigate and to recover safety function loss by applying and improving the resilience index.
Kasahara, Naoto*; Demachi, Kazuyuki*; Sato, Takuya*; Ichimiya, Masakazu*; Wakai, Takashi; Yamano, Hidemasa; Nakamura, Izumi*
no journal, ,
The conventional purpose in the field of structural strength has been to prevent damage to design basis events (DBE). For beyond design basis events (BDBE), it is necessary to mitigate the impact on safety on the premise that damage will occur. The authors propose a mitigation method that suppresses the consequence into a fracture mode with a large impact by reducing the load due to a fracture with a small impact on safety. We will introduce the research results for individual component, extend the applicable area to systems of components, and propose a new approach that contributes to improving plant safety.
Shobu, Takahisa; Tominaga, Aki; Sato, Seiichi*; Hatakeyama, Kiyoshi*; Nagai, Takayuki
no journal, ,
In order to ensure the long-term safety of the vitrified radioactive waste, the relationship between number of stirring and characteristics of material strength after melting the raw material glass and simulated vitrified radioactive waste were investigated by synchrotron radiation X-rays. As a result, it was clarified that the internal defects in simulated vitrified radioactive waste increased and the material strength of simulated vitrified radioactive waste was suitable as increasing the number of stirrings.
Sakaba, Nariaki
no journal, ,
The Japan Atomic Energy Agency (JAEA) has been developing High Temperature Gas-cooled Reactor technology since 1960s. In December 2020, Japanese Government clearly stated "Green Growth Strategy Through Achieving Carbon Neutrality in 2050" that noted a commitment to the milestone and developing and implementing significant efforts in various sectors for achieving carbon neutrality around 2050. High Temperature Gas-cooled Reactor (HTGR) is expected to play one of a dominant roles to reduce carbon generated from non-electric fields by utilizing heat and hydrogen produced by HTGR. To produce hydrogen from HTGR, it is necessary to establish a connecting technology including safety case between HTGR and hydrogen production process. The milestone for hydrogen production by HTTR (High Temperature Engineering Test Reactor) is given by the Green Growth Strategy. JAEA is now planning its R&D towards generation of hydrogen using heat from HTTR by 2030. This paper describes current status of HTGR R&D in JAEA.
Tamaoki, Yuichi; Isozaki, Ryosuke; Suzuki, Ryuta; Akada, Masataka; Sawahata, Satoshi; Yonezawa, Ryoma; Suzuki, Hisashi; Mizukoshi, Yasutaka; Sakamoto, Naoki
no journal, ,
The five post-irradiation examination facilities in the Oarai Research and Development Institute of the Japan Atomic Energy Agency have been operated for over 40 years in order to investigate the irradiation performance and physicochemical characteristics of nuclear fuels and materials for fast reactors. The equipments associated with these facilities have been managed to maintain secure from the problems that occurred in the process of aging. Therefore, we established a safety assessment method for aging facilities in 2002, and we have been conducting maintenance management of facilities since then. In this study, designing risk assessment flow is considered in order to solve the issues detected as a result of analysis based on the experience of the repairment during the periodic safety review monitoring.
Suzuki, Ryuta; Mizukoshi, Yasutaka
no journal, ,
no abstracts in English
Isozaki, Ryosuke; Sakamoto, Naoki
no journal, ,
no abstracts in English