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Takano, Kazuya; Sakamoto, Yoshihiko; Morohoshi, Kyoichi*; Okazaki, Hitoshi*; Gima, Hiromichi*; Teramae, Takuma*; Ikarimoto, Iwao*; Botte, F.*; Dirat, J.-F.*; Dechelette, F.*
Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 8 Pages, 2019/05
ASTRID has the objective to integrate innovative options in order to prepare the 4th generation reactors. In ASTRID, large number of tubes are installed above each fuel subassembly to monitor the core. These instrumentations such as thermocouples (TC) and Failed Fuel Detection and Location (FFDL) systems are integrated into Above Core Structure (ACS) with various sizes tubes. In the present study, the routing study for TC tubes and FFDL tubes was performed with 3D modeling and mock-up experiment of the ACS designed for ASTRID with 1500 MW thermal power in order to clarify the integration process and secure the design hypotheses. Although some problems on fabricability were found in the mock-up experiment, the possible solutions were proposed. The present study gives manufacturing feedback to design team and will contribute to increase the knowledge for ACS design and fabricability.
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.
Kawasaki, Nobuchika; Sakamoto, Yoshihiko; Eto, Masao*; Taniguchi, Yoshihiro*; Kamishima, Yoshio*
Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.760 - 769, 2015/05
The Japan Sodium-cooled Fast Reactor, JSFR, is currently under conceptual study. The concept of JSFR's reactor system is a compact reactor system to avoid excessive increase of reactor vessel diameter with structural and fluid integrities. To realize this concept, single rotating plug with advanced refueling system is adopted. Advanced refueling system consists of column type Upper Internal Structure and pantograph type Fuel Handling Machine. To realize structural and fluid integrities, top entry piping, sodium dam and flow block/guide structures are adopted. Structural integrities against seismic displacement or thermal stress and fluid integrities against vortex cavitations or cover gas entrainment can be ensured with these designs.
turbine system, 2; Turbine system and plant sizeKisohara, Naoyuki; Sakamoto, Yoshihiko; Kotake, Shoji*
JAEA-Research 2014-016, 60 Pages, 2014/09
JAEA-Research-2014-016.pdf:22.38MB
JAEA has performed a design study of an S-CO gas turbine system applied to the JSFR. In this study, the S-CO cycle turbine system was directly connected to the primary sodium system of the JSFR to eliminate the secondary sodium circuit, aiming for further economical improvement. This report describes the system configuration, heat/mass balance, and main components of the S-CO turbine system, based on the JSFR specifications. The layout of components and piping in the reactor and turbine buildings were examined and the dimensions of the buildings were estimated. The study has revealed that the reactor and turbine buildings could be reduced by 7% and 40%, respectively, in comparison with those in the existing JSFR design with the secondary sodium circuit employing the steam turbine. The cycle thermal was also calculated as 41.9-42.3%, which is nearly the same as that of the JSFR with the water/steam system.
turbine system, 1; Sodium/CO heat exchangerKisohara, Naoyuki; Sakamoto, Yoshihiko; Kotake, Shoji*
JAEA-Research 2014-015, 33 Pages, 2014/09
JAEA-Research-2014-015.pdf:27.33MB
JAEA has performed a design study of an S-CO gas turbine system applied to the JSFR. In this study, the S-CO cycle turbine system was directly connected to the primary sodium system of the JSFR to eliminate the secondary sodium circuit, aiming for further economical improvement. The Na/CO heat exchanger is one of the key components, and this report describes its structure and the safety in case of CO leak. A Printed Circuit Heat Exchanger (PCHE) is employed to the heat exchanger. A SiC/SiC ceramic composite material is used for the PCHE to prevent crack growth and to reduce thermal stress. The Na/CO heat exchanger has been designed in such a way that a number of small heat transfer modules are combined in the vessel in consideration of manufacture and repair. CO leak events in the heat exchanger have been also evaluated, and it revealed that no significant effect has arisen on the core or the primary sodium boundary.
Sakamoto, Yoshihiko; Garnier, J.-C.*; Rouault, J.*; Grandy, C.*; Fanning, T.*; Hill, R.*; Chikazawa, Yoshitaka; Kotake, Shoji*
Nuclear Engineering and Design, 254, p.194 - 217, 2013/01
Times Cited Count：16 Percentile：76.68(Nuclear Science & Technology)This trilateral study confirms that the fundamental mission of the fast reactor is to achieve significant uranium utilization and waste management goals. Thus, fast spectrum reactor concepts are vital for nuclear fuel cycle sustainability goals. The trilateral countries agree that SFR, GFR and LFR are capable to achieve the goals. However SFR is the most matured technology from the view point of industrial deployment while GFR and LFR still require long term development before a test reactor project could begin. The trilateral common view supports fast reactor project situations in each country.
Aoto, Kazumi; Uto, Nariaki; Sakamoto, Yoshihiko; Ito, Takaya*; Toda, Mikio*; Kotake, Shoji*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 11 Pages, 2012/00
In the FaCT project, SFR with 1,500 MWe is a target for the commercialization. R&D on innovative technologies to achieve the economic competitiveness and enhance the reliability and safety is carried out. A compact RV without wall-cooling layer is pursued in consideration of seismic reliability. For a two-loop cooling system with shortened high chromium steel piping, studies on the hydraulics in the pipe elbow and the fabrication capability of the pipes are being performed. A double-walled straight tube SG is investigated to enhance the reliability against sodium/water reaction, and developmental works are progressing including the thermal-hydraulic design and trial manufacturing for components. SASS is being developed with safety analysis of the applicability for JSFR and experimental demonstration in JOYO. An advanced fuel handling system is also pursued. Discussion on whether the innovative technologies can be adopted for JSFR is in progress to be finalized in 2010.
Eto, Masao*; Kamishima, Yoshio*; Okamura, Shigeki*; Watanabe, Osamu*; Oyama, Kazuhiro*; Negishi, Kazuo; Kotake, Shoji*; Sakamoto, Yoshihiko; Kamide, Hideki
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 10 Pages, 2012/00
In the JSFR design, the diameter of the Reactor Vessel (RV) shall be minimized and the reactor internal structures shall be simplified for reduction in construction cost. The reduction in the RV diameter is achieved by adopting an advanced refueling system and the hot RV with high temperature wall. The flow velocity in the reactor upper plenum increases because the diameter of the RV is decreased. As the result, the coolant flow field in reactor upper plenum is severe. The optimization of the coolant flow field in the reactor upper plenum was carried out for prevention the cover gas entrainment and the vortex cavitations at the hot leg intake. In addition, structural integrities for seismic loadings and thermal loadings were evaluated because the design seismic loading was highly increased and the vessel wall is directly exposed to the thermal transients of the upper plenum. This paper describes the characteristics and the results of the design study of the reactor system.
Okamura, Shigeki*; Eto, Masao*; Kamishima, Yoshio*; Negishi, Kazuo; Sakamoto, Yoshihiko; Kitamura, Seiji; Kotake, Shoji*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 10 Pages, 2012/00
This paper describes the seismic design of JSFR, which includes the seismic condition, the seismic isolation system and the seismic evaluation of primary component. JSFR employs a seismic isolation system to mitigate the earthquake force. The design seismic loading is made more severe than ever since Niigata-ken Chuetsu-oki Earthquake in 2007. The earthquake force loaded on the primary components has to be mitigated more than that of the previous seismic isolation system. We examined the advanced seismic isolation system by optimizing the performance of the previous seismic isolation system considering the natural frequency of the primary components. The advanced seismic isolation system for SFR was adopted laminated rubber bearings which are thicker than that of the previous, as well as oil dampers. The seismic evaluation of nuclear reactor components under applying the advanced seismic isolation system was performed; the performance of the system was confirmed.
Kawasaki, Nobuchika; Okamura, Shigeki*; Sawa, Naoki*; Sakamoto, Yoshihiko; Negishi, Kazuo
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 7 Pages, 2011/10
Japan Sodium-Cooled Fast Reactor adopts an compacted hot reactor vessel concept. From the point of structural designs to ensure both seismic design and elevated temperature design is important. In this study, based on a common conservative seismic loading condition considered with the Niigata-ken Chuetsu-oki Earthquake, seismic evaluations were carried out, the thicknesses of reactor vessels of 750 MWe and 500 MWe output plants were determined. For both plants 50 mm was selected as the thickness, and ensured buckling evaluation margins were more than 2.4. From the point of seismic design, the difference of plant output was negligible. With the condition of 50 mm thickness of reactor vessel, thermal integrities were evaluated. For three plant start-up conditions which were 2.2, 3.2, and 4.3 days, thermal ratcheting and creep-fatigue damage were evaluated. As a result plant start-up period needed more than 3.2 days for both 750 MWe and 500 MWe output plants. Caused thermal stress were the nearly same for both plants, therefore from the point of thermal design, the difference of plant output was negligible.
Aoto, Kazumi; Uto, Nariaki; Sakamoto, Yoshihiko; Ito, Takaya*; Toda, Mikio*; Kotake, Shoji*
Journal of Nuclear Science and Technology, 48(4), p.463 - 471, 2011/04
In the FaCT project, SFR with 1,500 MWe is a target for the commercialization. R&D on innovative technologies to achieve the economic competitiveness and enhance the reliability and safety is carried out. A compact RV without wall-cooling layer is pursued in consideration of seismic reliability. For a two-loop cooling system with shortened high chromium steel piping, studies on the hydraulics in the pipe elbow and the fabrication capability of the pipes are being performed. A double-walled straight tube SG is investigated to enhance the reliability against sodium/water reaction, and developmental works are progressing including the thermal-hydraulic design and trial manufacturing for components. SASS is being developed with safety analysis of the applicability for JSFR and experimental demonstration in JOYO. An advanced fuel handling system is also pursued. Discussion on whether the innovative technologies can be adopted for JSFR is in progress to be finalized in 2010.
Kisohara, Naoyuki; Sakamoto, Yoshihiko
JAEA-Review 2010-036, 26 Pages, 2010/09
JAEA-Review-2010-036.pdf:3.05MB
The prototype fast breeder reactor MONJU has been shut down since the secondary sodium leak accident that occurred in December 1995. After the accident, an investigation into the cause and a comprehensive safety review of the plant were conducted, and various countermeasures for sodium leak were examined. Modification work commenced in September 2005. Since sodium is used as coolant in MONJU, the modification work required work methods suitable for the handling of sodium. From this perspective, the use of a plastic bag when opening the sodium boundary, oxygen concentration control in a plastic bag, slightly-positive pressure control of cover gas in the systems, pressing and cutting with a roller cutter to prevent the incorporation of metal fillings, etc. were adopted. Owing to these work methods, the modification work proceeded close to schedule without incident.
Sakamoto, Yoshihiko; Kotake, Shoji; Aoto, Kazumi; Toda, Mikio*
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 9 Pages, 2010/05
JAEA is now performing a FaCT project. The first milestone is in 2010; decisions about whether or not to adopt innovative technologies in the JSFR design will be made in the year. Preliminary assessment is underway to produce recommendations for final discussion. This paper describes some important progress in the preliminary assessment. As for the reactor system design, structural integrity against both thermal stress and seismic force was investigated. Then, the specification of the reactor system was established. Also, investigation of design options to extend a design margin against seismic force has been suggested. Regarding thermal hydraulics issues, design measures have been introduced to restrain cover gas entrainment and vortex cavitations. Further investigation is now in progress for design optimization or improvement of preventive effect. Concerning the piping design of primary cooling circuit, the creep strength reduction by Type-IV damage was taken into account.
Kotake, Shoji; Sakamoto, Yoshihiko; Mihara, Takatsugu; Kubo, Shigenobu*; Uto, Nariaki; Kamishima, Yoshio*; Aoto, Kazumi; Toda, Mikio*
Nuclear Technology, 170(1), p.133 - 147, 2010/04
Times Cited Count：36 Percentile：91.14(Nuclear Science & Technology)Japan Atomic Energy Agency (JAEA) is now executing "Fast Reactor Cycle Technology Development (FaCT)" project in cooperation with the Japanese electric utilities. In the FaCT project, both the conceptual design study for Japan Sodium-cooled Fast Reactor (JSFR) and the developments of innovative technologies to be adopted to JSFR are now implemented with paying attention to the consistency between the design and the innovative technologies. The current target is that the development will be accomplished around 2015, after that a licensing procedure for the demonstration JSFR will be launched. This paper describes design requirements, design characteristics of JSFR and evaluation on the performances for economic competitiveness. Furthermore, the current status of the key technology development for JSFR is briefly introduced.
gas turbine power generation fast reactor(Contract research, Translated document)Kisohara, Naoyuki; Kotake, Shoji; Sakamoto, Yoshihiko
JAEA-Review 2008-040, 67 Pages, 2008/08
JAEA-Review-2008-040.pdf:3.93MB
(1) Preliminary design of an SFR that adopts a super-critical CO turbine has been made. This SFR system eliminates secondary sodium circuits. The power generation efficiency of the SFR is 42%. Compared to a SFR that adopts a steam cycle with secondary sodium circuits, the reactor building volume of the SC-CO SFR is reduced by 20%. (2) A super-critical CO cycle test loop was fabricated. The high efficiency of a compressor is confirmed near the super-critical region. The temperature efficiencies of PCHE recuperators are 98-99% (hot leg). No flow instability is observed in the loop operation. (3) Na/CO Reaction tests were executed. Continuous reaction occurs more than 570-580
C. Reaction products are NaCO
and CO. The reaction heat is 50-75kJ/Na-mol. (4) Safety calculation was done for 1 DEG tube failure in Na/CO heat exchanger. The maximum pressure in the primary circuit is below the allowed level. The void reactivity of the reactor core also has no affect. The reaction product brought no sodium flow blockage in fuel assemblies. (5) After an exposure of 600C-5000 hours in super-critical CO environment, the corrosion of 12Cr steel and 316FR were 170g/m
and 5g/m, respectively. 316 FR shows a good corrosion proof property.
Sakamoto, Yoshihiko; Kubo, Shigenobu*; Kotake, Shoji; Kamishima, Yoshio*
Dai-13-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.505 - 506, 2008/06
This paper describes the enhancement of reliability for components of the reactor system in JSFR design. As for manufacturability, compact design of the RV enables its manufacture in a factory. This results in high quality welding and in precise machining of the RV. The adoption of ring-shaped forgings contributes for securing the reliability against thermal stress as well as securing the dimension precision. Regarding maintenability, the in-vessel structures have simple configurations, so it is comparatively easy for inspection equipments to reach inspection targets. In the JSFR design, sodium boundary area is reduced significantly, which makes double-walled design of the piping easier, and reduces welding lines. So, the reactor system of JSFR is desirable to inspect the in-vessel structures efficiently, and there is a prospect of reliable plant operation. Advanced inspection technologies are also under development for the inspection of the in-vessel structures under sodium.
Sakamoto, Yoshihiko; Kubo, Shigenobu; Kotake, Shoji; Kamishima, Yoshio*
Proceedings of 2008 International Congress on Advances in Nuclear Power Plants (ICAPP '08) (CD-ROM), p.505 - 511, 2008/06
JSFR has the advanced loop type layout. In this paper, advantages of the advanced loop type reactor are presented in terms of reliability on the reactor structure. Compact design of the RV enables its manufacture in a factory which has high quality welding and precise machining. The RV has high reliability against thermal stress due to application of ring-shaped forgings around high stress parts. The in-vessel structures are simple, and this makes the approach to inspection targets easy. In JSFR, sodium boundary area is reduced significantly, which makes double-walled piping design easier, and reduces welding lines. So, the reactor structure of JSFR is desirable to inspect the in-vessel structures efficiently, and there is a prospect of reliable plant operation. Advanced technologies are also under development to inspect the structures immersed in sodium. In addition, the loop type reactor is suitable under severe earthquake condition as a result of comparative evaluation.
Sakamoto, Yoshihiko; Kotake, Shoji; Nishikawa, Akira; Enuma, Yasuhiro; Ando, Masato; Sagayama, Yutaka
Proceedings of 13th International Conference on Nuclear Engineering (ICONE-13), 0 Pages, 2005/05
The Feasibility Study on Commercialized Fast Reactor (FR) Cycle Systems is under way in order to propose prominent FR cycle systems that will respond to the diverse needs of society in the future. The design studies on various FR system concepts have been achieved and then the evaluations of potential to achieve the development targets have been carried out. Crucial issues have been found out for each FR system concept and their development plans for the key technologies are summarized as the roadmap. The characteristics and the differences in performances have been investigated. The crucial issues and the development periods have been clarified. Further investigation is now in progress. The promising concept will be proposed based on result of comparative evaluation at the end of the Phase II study.
Kotake, Shoji; Sakamoto, Yoshihiko; Enuma, Yasuhiro; Ando, Masato; Nishikawa, Akira; Sagayama, Yutaka
Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05), 0 Pages, 2005/05
The Feasibility Study on Commercialized Fast Reactor (FR) Cycle Systems is under way in order to propose prominent FR cycle systems that will respond to the diverse needs of society in the future. The design studies on various FR system concepts have been achieved and then the evaluations of potential to achieve the development targets have been carried out. Crucial issues have been found out for each FR system concept and their development plans for the key technologies are summarized as the roadmap. The characteristics and the differences in performances have been investigated. The crucial issues and the development periods have been clarified. Further investigation is now in progress. The promising concept will be proposed based on result of comparative evaluation at the end of the Phase II study.
Sagayama, Yutaka; Enuma, Yasuhiro; Sakamoto, Yoshihiko; Kotake, Shoji
Paper No.103, 0 Pages, 2004/11
Focusing on the cover layer materials (as the Radon Barrier Materials), which could have the effect to restrain the radon from scattering into the air and the effect of the radiation shielding, we produced the radon barrier materials with crude bentonite on an experimental basis, using the rotary type comprehensive unit for grinding and mixing, through which we carried out the evaluation of the characteristics thereof.
