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Nishida, Akemi
Doboku Gakkai Dai-14-Kai Kozobutsu No Shogeki Mondai Ni Kansuru Shinpojiumu Rombunshu (Internet), 5 Pages, 2024/01
no abstracts in English
Yamano, Hidemasa; Kurisaka, Kenichi; Takano, Kazuya; Kikuchi, Shin; Kondo, Toshiki; Umeda, Ryota; Shirakura, Shota*
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09
This project studies investigation on safety design guideline and risk assessment technology for sodium-cooled fast reactor with the molten-salt heat storage system, development of evaluation method for heat transferring performance between sodium and molten-salt and improvement of the performance, and evaluation of chemical reaction characteristic between sodium and molten-salt and improvement of its safety. The project overview is presented in this report.
Okuda, Takahiro; Takahashi, Hideki*; Watakabe, Tomoyoshi
Mechanical Engineering Journal (Internet), 10(4), p.23-00075_1 - 23-00075_9, 2023/08
In recent years, to make the seismic design more rational for the piping systems in nuclear power plants, it has been expected to develop a design method considering plastic deformation and the accompanying energy dissipation of the piping itself. In this study, an extensive series of seismic response analyses was conducted to investigate the degree of influence of the plastic deformation of the pipe support structures on the seismic response of the entire piping system. The analyses include; plasticity is considered for (1) none, (2) the piping only, (3) the support structure only, and (4) both the piping and the support structure.
Doda, Norihiro; Kato, Shinya; Hamase, Erina; Kuwagaki, Kazuki; Kikuchi, Norihiro; Ohgama, Kazuya; Yoshimura, Kazuo; Yoshikawa, Ryuji; Yokoyama, Kenji; Uwaba, Tomoyuki; et al.
Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.946 - 959, 2023/08
An innovative design system named ARKADIA is being developed to realize the design of advanced nuclear reactors as safe, economical, and sustainable carbon-free energy sources. This paper focuses on ARKADIA-Design for design studies as a part of ARKADIA and introduces representative verification methods for numerical analysis methods of the core design. ARKADIA-Design performs core performance analysis of sodium-cooled fast reactors using a multiphysics approach that combines neutronics, thermal-hydraulics, core mechanics, and fuel pin behavior analysis codes. To confirm the validity of these analysis codes, validation matrices are identified with reference to experimental data and reliable numerical analysis results. The analysis models in these codes and the representative practices for the validation matrices are described.
Doda, Norihiro; Nakamine, Yoshiaki*; Kuwagaki, Kazuki; Hamase, Erina; Kikuchi, Norihiro; Yoshimura, Kazuo; Matsushita, Kentaro; Tanaka, Masaaki
Keisan Kogaku Koenkai Rombunshu (CD-ROM), 28, 5 Pages, 2023/05
As a part of the development of the "Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle (ARKADIA)" to automatically optimize the life cycle of innovative nuclear reactors including fast reactors, ARKADIA-design is being developed to support the optimization of fast reactor in the conceptual design stage. ARKADIA-Design consists of three systems (Virtual plant Life System (VLS), Evaluation assistance and Application System (EAS), and Knowledge Management System (KMS)). A design optimization framework controls the connection between the three systems through the interfaces in each system. This paper reports on the development of an optimization analysis control function that performs design optimization analysis combining plant behavior analysis by VLS and optimization study by EAS.
Aoki, Takeshi; Shimizu, Atsushi; Noguchi, Hiroki; Kurahayashi, Kaoru; Yasuda, Takanori; Nomoto, Yasunobu; Iigaki, Kazuhiko; Sato, Hiroyuki; Sakaba, Nariaki
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 9 Pages, 2023/05
The safety design philosophy is developed for the HTTR (High Temperature Engineering Test Reactor) heat application test facility connecting high temperature gas-cooled reactor (HTGR) and the hydrogen production plant. The philosophy was proposed to apply proven conventional chemical plant standards to the hydrogen production facility for ensuring public safety against anticipated disasters caused by high pressure and combustible gases. The present study also proposed the safety design philosophy to meet specific safety requirements identified to the nuclear facilities with coupling to the hydrogen production facility such as measures to ensure a capability of normal operation of the nuclear facility against a fire and/or explosion of leaked combustible material, and fluctuation of amount of heat removal occurred in the hydrogen production plant. The safety design philosophy will be utilized to establish its basic and detailed designs of the HTTR-heat application test facility.
Yamano, Hidemasa; Kurisaka, Kenichi; Takano, Kazuya; Kikuchi, Shin; Kondo, Toshiki; Umeda, Ryota; Shirakura, Shota*; Hayashi, Masaaki*
Proceedings of 8th International Conference on New Energy and Future Energy Systems (NEFES 2023) (Internet), p.27 - 34, 2023/00
Times Cited Count：0This project studies investigation on safety design guideline and risk assessment technology for sodium-cooled fast reactor with the molten-salt heat storage system, development of evaluation method for heat transferring performance between sodium and molten-salt and improvement of the performance, and evaluation of chemical reaction characteristic between sodium and molten-salt and improvement of its safety. The project overview is presented in this report.
Geological Disposal Research and Development Department
JAEA-Evaluation 2022-007, 81 Pages, 2022/11
JAEA-Evaluation-2022-007.pdf:2.06MB
JAEA-Evaluation-2022-007-appendix(CD-ROM).zip:37.06MB
Japan Atomic Energy Agency (JAEA) consulted the advisory committee, "Evaluation Committee on Research and Development (R&D) Activities for Geological Disposal of High-Level Radioactive Waste", for post- and pre-review assessment of R&D activities on high-level radioactive waste disposal in accordance with "General Guideline for the Evaluation of Government Research and Development (R&D) Activities" by the Cabinet Office, Government of Japan, "Guideline for Evaluation of R&D in Ministry of Education, Culture, Sports, Science and Technology" and JAEA's "Regulation on Conduct for Evaluation of R&D Activities". In response to JAEA's request, the Committee reviewed mainly the progress of the R&D project on geological disposal, the relevance of the project outcome and the efficiency of the project implementation during the period of the current and next plan. This report summarizes the results of the assessment by the Committee with the Committee report attached.
Yamano, Hidemasa; Kubo, Shigenobu; Tokizaki, Minako*; Nakamura, Hironori*
Proceedings of International Conference on Topical Issues in Nuclear Installation Safety; Strengthening Safety of Evolutionary and Innovative Reactor Designs (TIC 2022) (Internet), 12 Pages, 2022/10
Specific design features of advanced sodium-cooled fast reactors (SFRs) designed in Japan are a passive reactor shutdown system, a passive decay heat removal system (DHRS), and an in-vessel retention (IVR) concept against an anticipated transients without scram (ATWS) in design extension condition (DECs). The present paper describes numerical analysis methodologies for event sequences studied in Japan and some numerical analyses of DECs to show the effectiveness of the passive shutdown system against a typical ATWS and severe accident mitigation measures for the IVR of molten core. For the passive shutdown capability, the numerical analysis has demonstrated the effectiveness of a self-actuated shutdown system against a severe ATWS event, for which the temperature response time was separately evaluated by a computational fluid dynamics (CFD) code. A recently developed debris-bed cooling analysis methodology coupled with a CFD code and a debris-bed module has successfully simulated a three-dimensional coolant flow field near the debris bed with the passive DHRS capability in order to demonstrate the debris-bed coolability on a core catcher.
Takayanagi, Tomohiro; Sugita, Moe; Ueno, Tomoaki*; Horino, Koki*; Ono, Ayato; Yamamoto, Kazami; Kinsho, Michikazu
IEEE Transactions on Applied Superconductivity, 32(6), p.4101405_1 - 4101405_5, 2022/09
Times Cited Count：0 Percentile：0(Engineering, Electrical & Electronic)In order to design a pulsed electromagnet, it is necessary to consider eddy currents that depend on the pattern shape of the pulse excitation, and to design the structure and set the operating parameters considering the fluctuation of the magnetic field distribution and its effect on heat generation. Evaluation tests of a new bump magnet for the J-PARC RCS showed that the magnetic field distributions at the rising edge of the trapezoidal pattern and at the flat top are different from each other. This was also confirmed by the 3D dynamic magnetic field analysis of OPERA-3d. Since the skin-depth due to eddy currents depends on the time variation of the waveform, the effect is the same as changing the shape of the coil. The measurement of the magnetic field distribution fluctuation was verified and evaluated using a flux meter, search coil, and Hall probe with different measurement principles.
Uchibori, Akihiro; Sogabe, Joji; Okano, Yasushi; Takata, Takashi*; Doda, Norihiro; Tanaka, Masaaki; Enuma, Yasuhiro; Wakai, Takashi; Asayama, Tai; Ohshima, Hiroyuki
Proceedings of Technical Meeting on State-of-the-art Thermal Hydraulics of Fast Reactors (Internet), 10 Pages, 2022/09
The ARKAIDA has been developed to realize automatic optimization of plant design from safety evaluation for the advanced reactors represented by a sodium-cooled fast reactor. ARKADIA-Design offers functions to support design optimization both in normal operating conditions and design basis events. The multi-level simulation approach by the coupled analysis such as neutronics, core deformation, core thermal hydraulics was developed as one of the main technologies of the ARKADIA-Design. On the other hand, ARKAIDA-Safety aims for safety evaluation considering severe accidents. As a key technology, the numerical methods for in- and ex-vessel coupled phenomena during severe accidents in sodium-cooled fast reactors were tested through a hypothetical severe accident event.
Ogawa, Rina; Nakata, Hisakazu; Sugaya, Toshikatsu; Sakai, Akihiro
JAEA-Technology 2022-010, 54 Pages, 2022/07
JAEA-Technology-2022-010.pdf:11.07MB
Japan Atomic Energy Agency has considered trench disposal as one of the disposal methods for radioactive wastes generated from research facilities and other facilities. The trench disposal facility is regulated by "Act on the Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors". In particular, the design of the trench facility is regulated by a rule under the law. When the rule was amended in 2019, the design of the trench disposal facility required equipment to reduce ingress of rain water and groundwater. In the report, studies on the design of a trench disposal facility to adapt to the amended rule were performed. The trench disposal facility has considered being established in a place lower than groundwater level. Therefore, it was decided to study covering soil at the upper part of the trench facility, because the ingress water in the facility is mainly derived from rain water. In this study, it was decided to evaluate the design of covering soil of the radioactive waste categorized into chemically stable materials. Therefore, as the examination method, a parameter study on varying the permeability coefficient and thickness of the layers composing cover soil. In the parameter study, the velocity of the water infiltrating into the trench facility was evaluated. Based on the results, more efficient design of the layers composing the covering soil was considered. The result showed that the impermeable efficiency of the covering soil was different depending on the thickness and the permeability conductivity of each layer. As a result, it was possible to understand the impermeable performance of covering soil by the permeability coefficient and thickness of each layer. We will plan to decide the specification of the cover soil while examination of future tasks and cost in the basic design.
Tanaka, Masaaki; Doda, Norihiro; Yokoyama, Kenji; Mori, Takero; Okajima, Satoshi; Hashidate, Ryuta; Yada, Hiroki; Oki, Shigeo; Miyazaki, Masashi; Takaya, Shigeru
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2022/07
To assist conceptual studies of various reactor systems conducted by private sectors in nuclear power innovation, development of an innovative design system named ARKADIA (Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle) is undergoing to achieve the design of an advanced nuclear reactor as a safe, economic, and sustainable carbon-free energy source. In this paper, focusing on the ARKADIA-Design as a part of it, the progress in the development of optimization processes on the representative problems in the fields of the core design, the plant structure design, and the maintenance schedule planning are introduced.
Hazama, Taira
Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, p.87 - 161, 2022/07
Ohgama, Kazuya; Hara, Toshiharu*; Ota, Hirokazu*; Naganuma, Masayuki; Oki, Shigeo; Iizuka, Masatoshi*
Journal of Nuclear Science and Technology, 59(6), p.735 - 747, 2022/06
Times Cited Count：1 Percentile：31.61(Nuclear Science & Technology)Doda, Norihiro; Nakamine, Yoshiaki*; Igawa, Kenichi*; Iwasaki, Takashi*; Murakami, Satoshi*; Tanaka, Masaaki
Keisan Kogaku Koenkai Rombunshu (CD-ROM), 27, 6 Pages, 2022/06
As a part of the development of the "Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle (ARKADIA)" to automatically optimize the life cycle of innovative nuclear reactors including fast reactors, ARKADIA-design is being developed to support the optimization of fast reactor design in the conceptual stage. ARKADIA-Design consists of three systems (Virtual plant Life System (VLS), Evaluation assistance and Application System (EAS), and Knowledge Management System (KMS)). A design optimization framework controls the cooperation between the three systems through the interfaces in each system. This paper reports on the development status of the "VLS interface," which has a control function of coupling analysis codes in VLS.
Tanaka, Masaaki; Doda, Norihiro; Mori, Takero; Yokoyama, Kenji; Uwaba, Tomoyuki; Okajima, Satoshi; Matsushita, Kentaro; Hashidate, Ryuta; Yada, Hiroki
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03
Japan Atomic Energy Agency is developing an innovative design system named ARKADIA to achieve the design of an advanced nuclear reactor as a safe, economic, and sustainable carbon-free energy source. In the first phase of its development, ARKADIA-Design for design study and ARKADIA-Safety for safety assessment will be developed individually. In this paper, focusing on the ARKADIA-Design, the concept of the system is described and numerical analysis codes to be used for the multi-level and multi-physics analyses are introduced. Descriptions of the practical functions composed by the analysis codes and the representative problems in application studies for validation are introduced.
Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Doda, Norihiro; Tanaka, Masaaki
Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 8 Pages, 2021/08
In the design study of an advanced sodium-cooled fast reactor (Advanced-SFR) in JAEA, the use of a specific fuel assembly (FA) with an inner duct structure called FAIDUS has been investigated to enhance safety of Advanced-SFR. Due to the asymmetric layout of fuel rods by the inner duct, it is necessary to estimate the temperature distribution to confirm feasibility of FAIDUS. For the FAIDUS, confirmation of validity of the numerical results by a subchannel analysis code named ASFRE was not enough because the reference data on the thermal hydraulics in FAIDUS have not been obtained by the mock-up experiment, yet. Therefore, the code-to-code comparisons with numerical results of ASFRE and those of a CFD code named SPIRAL was conducted. The applicability of ASFRE was indicated through the confirmation of the consistency of mechanism of the specific temperature and velocity distributions appearing around the inner duct between the numerical results by ASFRE and those by SPIRAL.
Ono, Ayato; Takayanagi, Tomohiro; Ueno, Tomoaki*; Horino, Koki*; Yamamoto, Kazami; Kinsho, Michikazu
JAEA-Technology 2020-023, 40 Pages, 2021/02
JAEA-Technology-2020-023.pdf:2.98MB
The 3 GeV rapid cycling synchrotron of Japan Proton Accelerator Research Complex (J-PARC) uses a large number of electromagnet power supplies in order to generate a high-intensity beam of 1 MW. These devices have been specially developed to meet the required specifications of the proton beams. Ten years have passed since the 3 GeV synchrotron had started operation, and we need to replace and update of the components due to failures caused by the aging deterioration. Since the J-PARC is used by many users, it is quite important to recover as soon as possible when a trouble occurs. However, we often spend lots of time to investigate the status and cause of the problem, then it results in the delay of recovery work. One of the major reasons is due to the differences in the manufacturers of sensors and monitors. Therefore, we have to create a manual for each power supply and prepare some exclusive tools. However, troubles rarely occur in the same state and situation, so we have to rely on the experience and knowledge. Even for power supplies with different purposes and specifications, some components, such as sensors, can be shared in many cases. In addition, if the concept of the interlock system, for monitoring the status of the power supply and detecting malfunctions, is shared between the different power supplies, the method and response for failure investigation can be standardized. By using a device with good maintainability, the accelerator operation will be more stable and reliable. In this report, we introduce the necessity of sharing the design concept and common parts. We also explain the basic design model for safety and reliability, using an example of manufacturing an electromagnet power supply for the 3 GeV synchrotron.
Kato, Masato
Encyclopedia of Nuclear Energy, Vol.2, p.298 - 307, 2021/00
