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Kamide, Hideki; Kawasaki, Nobuchika; Hayafune, Hiroki; Kubo, Shigenobu; Chikazawa, Yoshitaka; Maeda, Seiichiro; Sagayama, Yutaka; Nishihara, Tetsuo; Sumita, Junya; Shibata, Taiju; et al.
Jisedai Genshiro Ga Hiraku Atarashii Shijo; NSA/Commentaries, No.28, p.14 - 36, 2023/10
Developments of next generation nuclear reactors, e.g., Fast Reactor, and High Temperature Gas cooled Reactor, are in progress. They can contribute to markets of electricity and industrial heat utilization in the world including Japan. Here, current status of reactor developments in Japan and also situation in the world are summarized, especially for activities of Generation IV International Forum (GIF), developments of Fast Reactor and High Temperature Gas cooled Reactor in Japan, and SMR movements in the world.
Akimoto, Yuta; Ezure, Toshiki; Onojima, Takamitsu; Kurihara, Akikazu
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 9 Pages, 2023/05
A numerical analysis method has been developed to evaluate thermal-hydraulic behaviors in a reactor vessel under the operation of a NC-DHRS at the Japan Atomic Energy Agency. During the validation of the evaluation method, in addition to uncertainties due to the numerical solution and input parameters in simulations, it is important to quantify uncertainties due to the experimental data. From this perspective, JAEA has been developing an experimental database and uncertainty evaluation methods for sodium experiments during operation of the NC-DHRS. In this study, the authors have proposed an uncertainty evaluation approach during relative calibrations of thermocouples in sodium experiments. The proposed approach was applied to experimental data obtained in a sodium NC-DHRS experiment conducted at PLANDTL-2. As a result, uncertainties of the experimental data were successfully evaluated and the applicability of the method to temperature measurement in sodium experiments was confirmed.
Tanaka, Masaaki; Miyake, Yasuhiro*; Ezure, Toshiki; Hamase, Erina
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 9 Pages, 2023/05
The numerical analysis model for the computational fluid dynamics (CFD) code for the design study is developed to evaluate the thermal-hydraulics in the core under the core-plenum interaction (CPI) during the decay heat removal using the dipped type direct heat exchanger (D-DHX). To judge the adequacy of the numerical results for a validation study with the sodium experiment results conducted at PLANDTL-2 facility, the degree of difference (DoD) between the numerical and experimental results must be measured by using the area validation metrics (AVM). Through the examinations, the applicability of the AVM and MAVM based on the p-box method was confirmed.
Kubo, Shigenobu; Chikazawa, Yoshitaka; Ohshima, Hiroyuki; Kamide, Hideki
Handbook of Generation IV Nuclear Reactors, Second Edition, p.173 - 194, 2023/03
Handbook of Generation IV Nuclear Reactors, Second Edition is a fully revised and updated comprehensive resource on the latest research and advances in generation IV nuclear reactor concepts. Editor Igor Pioro and his team of expert contributors have updated every chapter to reflect advances in the field since the first edition published in 2016. JAEA contributes to Chapter 5; Sodium-cooled Fast Reactors (SFRs) and Chapter 12; Generation-IV Sodium-cooled Fast Reactor (SFR) concepts in Japan. Major characteristics and current technology developments including safety enhancement were described in Chapter 5. Chapter 12 shows design activities of SFR. Innovative technology developments, and update of the Japan sodium-cooled fast reactor design with lessons learned from the TEPCO Fukushima Daiichi NPP accident.
Kobayashi, Jun; Aizawa, Kosuke; Ezure, Toshiki; Nagasawa, Kazuyoshi*; Kurihara, Akikazu; Tanaka, Masaaki
JAEA-Research 2022-009, 125 Pages, 2023/01
JAEA-Research-2022-009.pdf:29.22MB
The design studies of an advanced loop-type sodium-cooled fast reactor (Advanced- SFR) have been carried out by the Japan Atomic Energy Agency (JAEA). At the core outlet, temperature fluctuations occur due to mixing of hot sodium from the fuel assembly with cold sodium from the control rod channels and radial blanket assembly. These temperature fluctuations may cause high cycle thermal fatigue around a bottom of Upper Internal Structure (UIS) located above the core. Therefore, we conducted a water experiment using a 1/3 scale 60 degree sector model that simulated the upper plenum of the advanced loop-type sodium-cooled reactor. And we proposed some countermeasures against large temperature fluctuations that occur at the bottom of the UIS. In this report, we have summarized that the effect of the countermeasure structure to mitigate the temperature fluctuation generated at the bottom of UIS is confirmed, and the Reynolds number dependency of the countermeasure structure and the characteristics of the temperature fluctuation on the control rod surface.
Kobayashi, Jun; Aizawa, Kosuke; Ezure, Toshiki; Kurihara, Akikazu; Tanaka, Masaaki
Hozengaku, 20(3), p.89 - 96, 2021/10
Hot sodium from the fuel assembly can mix with cold sodium from the control rod (CR) channel and the blanket assemblies at the bottom plate of the Upper Internal Structure (UIS) of Advanced-SFR. Temperature fluctuation due to mixing of the fluids at different temperature between the core outlet and cold channel may cause high cycle thermal fatigue on the structure around the bottom of UIS. A water experiment using a 1/3 scale 60 degree sector model simulating the upper plenum of the Advanced-SFR has been conducted to examine countermeasures for the significant temperature fluctuation generated around the bottom of UIS. We focused on the temperature fluctuations near the primary and backup control rod channels, and studied the countermeasure structure to mitigate the temperature fluctuation through temperature distribution and flow velocity distribution measurements. As a result, effectiveness of the countermeasure to mitigate the temperature fluctuation intensity was confirmed.
Kobayashi, Jun; Aizawa, Kosuke; Ezure, Toshiki; Kurihara, Akikazu; Tanaka, Masaaki
Hozengaku, 20(3), p.97 - 101, 2021/10
Focusing on the thermal striping phenomena that occurs at a bottom of the internal structure of an advanced sodium-cooled fast reactor (Advanced-SFR) that has been designed by the Japan Atomic Energy Agency, a water experiment using a 1/3 scale 60 degree sector model simulating the upper plenum of the Advanced-SFR has been conducted to examine countermeasures for the significant temperature fluctuation generated around the bottom of Upper Internal Structure (UIS). In the previous paper, we reported the effect of measures to mitigate temperature fluctuations around the control rod channels. In this paper, the same test section was used, and a water experiment was conducted to obtain the characteristics of temperature fluctuations around the radial blanket fuel assembly. And the shape of the Core Instrumentation Support Plate (CIP) was modified, and it was confirmed that it was highly effective in alleviating temperature fluctuations around the radial blanket fuel assembly.
Kamide, Hideki
Karyoku Genshiryoku Hatsuden, 71(11), p.638 - 648, 2020/11
Development status of next generation reactors is outlined mainly for sodium cooled fast reactor developed in Japan Atomic Energy Agency (JAEA). Development strategy in Japan, status of the advanced reactor developments including SMR over the world, and also results of research and development in JAEA are explained.
Negishi, Hitoshi; Kamide, Hideki; Maeda, Seiichiro; Nakamura, Hirofumi; Abe, Tomoyuki
Nihon Genshiryoku Gakkai-Shi ATOMO
, 62(8), p.438 - 441, 2020/08
Prototype Fast Breeder Reactor, Monju, was under decommission since April, 2018. It is the first time for Japan to make a sodium cooled reactor into decommission. It is significant work and will take 30 years. The Monju has provided wide spectrum and huge amount of findings and knowledge, e.g., design, R&D, manufacturing, construction, and operation up to 40% of full power over 50 years of development history. It is significant to utilize such findings and knowledge for the development and commercialization of a fast rector in Japan.
Ohshima, Hiroyuki; Kamide, Hideki
Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.2095 - 2107, 2019/08
Development of a sodium-cooled fast reactor has been implemented in Japan from the viewpoint of severe accident countermeasures. This paper describes the progress of research and development related to safety enhancement and the severe accident countermeasures. A volcanic PRA methodology was developed for the proper consideration of external hazards. Water and sodium experiments were carried out for the decay heat removal in a core disruptive accident (CDA), and also thermal hydraulic interactions between the core and upper plenum where dipped heat exchanger was operated. In order to elucidate the behavior of molten fuel during CDA, basic experiments of core melt fragmentation in deep and shallow sodium pools were carried out. X-ray visualization showed the liquid column of molten steel was intensively fragmented nearly simultaneously with a rapid expansion of sodium vapor.
Tsuruga Comprehensive Research and Development Center
JAEA-Technology 2019-007, 159 Pages, 2019/07
JAEA-Technology-2019-007.pdf:19.09MBJAEA-Technology-2019-007-high-resolution1.pdf:42.36MBJAEA-Technology-2019-007-high-resolution2.pdf:33.56MBJAEA-Technology-2019-007-high-resolution3.pdf:38.14MBJAEA-Technology-2019-007-high-resolution4.pdf:48.82MBJAEA-Technology-2019-007-high-resolution5.pdf:37.61MB
This report summarizes the history and achievements of the prototype fast breeder reactor Monju. The development of Monju started in 1968 as a prototype reactor following the experimental fast reactor Joyo. The development covers all the activity related to the fast reactor; plant design, mockup tests, construction, operation, and plant management. This report summarizes the history and achievements for 11 technical areas: history and principal achievements, design and construction, operation test, plant safety, core physics, fuel, plant system, sodium technology, materials and mechanical design, plant management, and trouble management.
Kamide, Hideki; Sakamoto, Yoshihiko; Kubo, Shigenobu; Oki, Shigeo; Ohshima, Hiroyuki; Kamiyama, Kenji
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06
Development of a sodium-cooled fast reactor has been implemented in Japan from the viewpoint of severe accident countermeasures in order to strengthen safety of a fast reactor since the Great East Japan Earthquake. This paper describes the progress of design study and research and development related to safety enhancement and the severe accident countermeasures. For the purpose of strengthening of decay heat removal function, several researches have been carried out on the decay heat removal in a core disruptive accident (CDA), diversity and applicability of decay heat removal systems, and thermal hydraulic evaluation methods. In order to elucidate the behavior of molten fuel during CDA, some in-pile and out-of-pile tests has been performed by international collaboration including basic experiments. Core design was also improved from the viewpoint of preventing the occurrence of severe accident.
Kamide, Hideki; Ohshima, Hiroyuki; Sakai, Takaaki; Tanaka, Masaaki
Nuclear Engineering and Design, 312, p.30 - 41, 2017/02
Times Cited Count：6 Percentile：51.46(Nuclear Science & Technology)In the framework of the Generation-IV International Forum, the safety design criteria (SDC) incorporating safety-related Research and Development results on innovative technologies and lessons learned from Fukushima Dai-ichi Nuclear Power Plants accident has been established to provide the set of general criteria for the safety designs of structures, systems and components of Generation-IV Sodium-cooled Fast Reactors (Gen-IV SFRs). A number of thermal-hydraulic evaluations are necessary to meet the concept of the criteria in the design studies of Gen-IV SFRs. This paper focuses on four kinds of thermal-hydraulic issues associated with the SDC, i.e., fuel subassembly thermal-hydraulics, natural circulation decay heat removal, core disruptive accidents, and thermal striping. Progress of evaluation methods on these issues is shown with activities on verification and validation (V and V) and experimental studies towards commercialization of SFR in Japan. These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all possible phenomena in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing our knowledge and technologies down.
Kobayashi, Jun; Ezure, Toshiki; Tanaka, Masaaki; Kamide, Hideki
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 5 Pages, 2016/11
JAEA has been conducting a design study for an advanced large-scale sodium-cooled fast reactor (SFR). Hot sodium from the fuel subassembly can mix with the cold sodium from the control rod (CR) channel at the bottom of Upper Internal Structure (UIS). Temperature fluctuation due to the fluid mixing at the core outlet may cause high cycle thermal fatigue at the bottom of UIS. JAEA had performed a water experiment to examine countermeasures for the significant temperature fluctuation generated at the bottom of SFRs UIS. Meanwhile, a self-actuated shutdown system (SASS) is equipped in a backup control rod (BCR) channel to ensure reactor shutdown. The BCR guide tubes have a flow guide structure "flow-collector" to provide reliable operation of SASS. Flow-collector may affect the thermal mixing behavior at the bottom of the UIS. This study has investigated the influence of the flow- collector on characteristics of the temperature fluctuation around the BCR channels.
Ono, Ayako; Kamide, Hideki; Kobayashi, Jun; Doda, Norihiro; Watanabe, Osamu*
Journal of Nuclear Science and Technology, 53(9), p.1385 - 1396, 2016/09
Times Cited Count：11 Percentile：71.62(Nuclear Science & Technology)Decay heat removal by natural circulation is a significant passive safety measure of a fast reactor against station blackout. The decay heat removal system (DHRS) of the loop type sodium fast reactor being designed in Japan comprises a direct reactor auxiliary cooling system and primary reactor auxiliary cooling system (PRACS). The thermal hydraulic phenomena in the plant under natural circulation conditions need to be understood for establishing a reliable natural circulation driven DHRS. In this study, sodium experiments were conducted using a plant dynamic test loop to understand the thermal-hydraulic phenomena considering natural circulation in the plant. The experiments simulating the scram transient confirmed that PRACS started up smoothly under natural circulation, and the simulated core was stably cooled after the scram. Moreover, the experiments varying the pressure loss coefficients of the loop as the experimental parameters showed robustness of the PRACS.
Ono, Ayako; Onojima, Takamitsu; Doda, Norihiro; Miyake, Yasuhiro*; Kamide, Hideki
Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.2183 - 2192, 2016/04
Some auxiliary cooling systems to remove the decay heat of the core are under consideration for a sodium-cooled fast reactor, and two of the typical systems are primary reactor auxiliary cooling system (PRACS) and direct reactor auxiliary cooling system (DRACS). In this study, sodium experiments were conducted in order to confirm the applicability of the PRACS and DRACS under a situation assuming the severe accidents with core melting. The plant dynamics test loop was used for these experiments, which contains a simulated core, the PRACS and DRACS. The core melt situation is simulated by shutting off the inlet of subassemblies (S/A). The experimental results revealed the cooling process of the partially/completely inactive S/A and confirmed the long-term heat removal by the PRACS/DRACS.
Kobayashi, Jun; Tanaka, Masaaki; Ohno, Shuji; Ohshima, Hiroyuki; Kamide, Hideki
Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.6664 - 6677, 2015/08
Numerical simulation is recognized an essential tool for the physical phenomena analysis and plant design study of a sodium-cooled fast reactor (SFR). In order to enhance credibility of the numerical results in the activities for plant design by using numerical simulations, it is recognized that verification and validation (V&V) process is very important. In this study, experiments for planar triple parallel jets mixing phenomena conducted in JAEA were proposed as benchmark problems for the code validation in the area of thermal striping study in the SFR development.
Kobayashi, Jun; Ezure, Toshiki; Kamide, Hideki; Oyama, Kazuhiro*; Watanabe, Osamu*
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05
A column type upper internal structure (UIS) is installed in the upper plenum of reactor vessel in JSFR. High cycle thermal fatigue may occur at the bottom plate (CIP) of the UIS where the hot sodium from the fuel subassembly can mix with the cold sodium from the control rod channel and the blanket fuel subassembly. We have been conducted a water experiment using a reactor upper plenum model to grasp the thermal-hydraulic phenomena around control rod (CR) channels, and to obtain countermeasures for significant temperature fluctuation on the CIP. The experimental apparatus has 1/3 scale and 60
sector model of the reactor upper plenum. By the experiment, characteristics of fluid temperature fluctuation between the handling head of the assemblies and the CIP are measured and countermeasure for the significant temperature fluctuation generation will be discussed on the influence of the distance from the handling head outlet to the lower surface of the CIP.
Kamide, Hideki; Ono, Ayako; Kimura, Nobuyuki; Endo, Junji*; Watanabe, Osamu*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13), Companion CD (CD-ROM), 11 Pages, 2015/04
Natural circulation decay heat removal is one of the significant issues for fast reactor safety, especially in long term station blackout events. Several sodium experiments were carried out using a 7-subassmbly core model for core thermal hydraulics under natural circulation conditions and for onset transients of natural circulation in a decay heat removal system (DHRS) including natural draft. Significant heat removal via inter-wrapper flow was confirmed in the experiments. Solidification of sodium in an air cooler is one of key issues in loss of heat sink events. Natural circulation characteristics under long-term decay heat removal were also obtained. Multi-dimensional phenomena, e.g., thermal stratification and bypass flow in plenums and/or heat exchangers, may influence the natural circulation. Thus, 3D simulation method was developed for entire region in the primary loop. Comparison of temperature distributions in a DHRS heat exchanger between experiment and analysis was done.
Hirata, Shingo; Chikazawa, Yoshitaka; Kato, Atsushi; Uto, Nariaki; Obata, Hiroyuki*; Kotake, Shoji*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 10 Pages, 2012/00
In Fast Reactor Cycle Technology Development (FaCT) project, Japan Sodium-cooled Fast Reactor (JSFR) is going to adopt an advanced fuel handling system. As for ex-vessel spent fuel handling, a pot which contains two fuel subassemblies simultaneously and is applicable size to compact reactor vessel, has been developing so as to shorten a refueling period leading to an improvement of plant availability. The pot is required to provide sufficient cooling capability even in case of transportation malfunction during transportation of spent fuel subassemblies with high decay heat. In the present study, experimental and analytical studies to evaluate the cooling capacity of the pot are summarized.
