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Yamano, Hidemasa; Takai, Toshihide; Emura, Yuki; Fukuyama, Hiroyuki*; Nishi, Tsuyoshi*; Morita, Koji*; Nakamura, Kinya*; Pellegrini, M.*
Nihon Kikai Gakkai 2023-Nendo Nenji Taikai Koen Rombunshu (Internet), 5 Pages, 2023/09
A research project has been conducting thermophysical property measurement of a eutectic melt, eutectic melting reaction and relocation experiments, eutectic reaction mechanism investigation, and physical model development on the eutectic melting reaction for reactor application analysis in order to simulate the eutectic melting reaction and relocation behavior of boron carbide as a control rod material and stainless steel during a core disruptive accident in an advanced sodium-cooled fast reactor designed in Japan. This paper describes the project overview and progress until JFY2022.
Kato, Atsushi; Yamamoto, Tomohiko; Ando, Masato; Chikazawa, Yoshitaka; Murakami, Hisatomo*; Oyama, Kazuhiro*; Kaneko, Fumiaki*; Higurashi, Koichi*; Chanteclair, F.*; Chenaud, M.-S.*; et al.
EPJ Nuclear Sciences & Technologies (Internet), 8, p.11_1 - 11_10, 2022/06
This paper provides an overview of plant system studies to establish a common technical view for Sodium-cooled Fast Reactor concept between France and Japan based on ASTRID600 and the new concept with downsized output called ASTRID150. One of important issues on a reactor structure design is to enhance seismic resistance to be tolerable against strong earthquake such that postulated in Japan. A concept of High Frequency Design is shared, and the design options related to HFD have been examined and design recommendations are established. In addition, this paper include results of studies for a steam generator, a decay heat removal system, a fuel handling system and a containment vessel.
Kato, Atsushi; Kubo, Shigenobu; Chikazawa, Yoshitaka; Miyagawa, Takayuki*; Uchita, Masato*; Suzuno, Tetsuji*; Endo, Junji*; Kubo, Koji*; Murakami, Hisatomo*; Uzawa, Masayuki*; et al.
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 11 Pages, 2022/04
The authors are carrying out conceptual design studies for a pool-type sodium-cooled fast reactor. There are main challenges such as measures against severe earthquake in Japan, thermal hydraulic in a reactor vessel (RV), a decay heat removal system design. When the JP-pool SFR of 650 MWe is installed in Japan, it shall be designed against the severe seismic conditions. Additionally, a newly three-dimensional seismic isolation system is under development.
Kudo, Hideyuki*; Inuzuka, Taisuke*; Hara, Masahide*; Kato, Atsushi; Nagai, Keiichi; Ide, Akihiro*
Journal of Nuclear Science and Technology, 57(1), p.9 - 23, 2020/01
Times Cited Count：1 Percentile：8.95(Nuclear Science & Technology)In sodium-cooled fast reactors (SFRs), it is necessary to remove the sodium remaining on spent fuel assemblies (FAs) before storing them in a spent fuel water pool (SFP) in order to minimize plant operating loads. A next-generation SFR in Japan has adopted an advanced dry cleaning process which consists of the following steps: argon gas blowing to remove the metallic residual sodium on the FA, moist argon gas blowing to deactivate the residual sodium, and direct storage in the SFP. This process increases economic competitiveness and reduces waste products. In this R&D work, performance of the dry cleaning process has been investigated. This paper describes experimental and analytical work focusing on the amount of residual sodium remaining on FA components, for instance the handling head, the wrapper tube, the upper shielding, and the entrance nozzle which was conducted after investigation of residual sodium on fuel pin bundles as a part of series study of the cleaning process.
Ide, Akihiro*; Kudo, Hideyuki*; Inuzuka, Taisuke*; Hara, Masahide*; Kato, Atsushi; Ishikawa, Nobuyuki; Otaka, Masahiko; Nagai, Keiichi; Saito, Junichi; Ara, Kuniaki
Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 10 Pages, 2019/05
A next generation SFR in Japan has adopted an advanced dry cleaning system which consists of the following process of argon gas blowing to reduce the amount of metallic sodium, moist argon gas blowing to deactivate the residual sodium, and direct storage in the SFP without using storage containers. This three-step process increases economic competitiveness and reduces waste products. In this Research and Development work, the amount of residual sodium and performance of the dry cleaning process were investigated. This paper describes experimental and analytical work for all parts of a fuel assembly except for a fuel pin bundle.
Chikazawa, Yoshitaka; Takaya, Shigeru; Tagawa, Akihiro; Kubo, Shigenobu
Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 6 Pages, 2019/05
A maintenance management required to prototype nuclear power reactors has been developed. One of important mission of a prototype reactor is to develop maintenance program for commercial reactors step by step securing safety. Since operating experience at the early stage is limited, the maintenance program for the prototype reactor should be a progressive one. It has to be modified and improved frequently taking into account R&D insight and operation experiences. Additionally, the maintenance program has to consider features of the prototype reactor even at the early stage. To select maintenance grades on particular components/systems, risk informed and graded approaches are effective. And maintenance programs have to take into account degradation mechanism originally due to reactor features. In this paper, applications for maintenance program on sodium valves of prototype fast breeder reactor Monju are studied as an example of prototype sodium-cooled reactors (SFR).
Kudo, Hideyuki*; Otani, Yuichi*; Hara, Masahide*; Kato, Atsushi; Ishikawa, Nobuyuki; Otaka, Masahiko; Nagai, Keiichi; Saito, Junichi; Ara, Kuniaki; Ide, Akihiro*
Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 10 Pages, 2019/05
A next generation SFR in Japan has adopted an advanced dry cleaning system which consists of the argon gas blowing process to reduce the amount of metallic residual sodium remaining on spent fuel assemblies. This paper describes experimental and analytical work focusing on the amount of residual sodium remaining on a fuel pin bundle before and after the argon gas blowing process. The experiments were conducted using a sodium test loop and a short specimen consisting of a 7 pin bundle. The effects of the blowing gas velocity and the blowing time were quantitatively analyzed in the experiments. On the basis of these experimental results, evaluation models predicting the amount of the residual sodium were constructed.
Hayafune, Hiroki; Glatz, J.-P.*; Yang, H.*; Ruggieri, J.-M.*; Kim, Y.-I.*; Ashurko, Y.*; Hill, R.*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 12 Pages, 2017/06
The SFR system arrangement Phase II became effective on 16 February 2016 by signatures of CEA, JAEA, KAERI, USDOE, and Rosatom, and was extended for additional 10 years. Collaboration of GIF SFR is growing adding new reactor concepts and related RDs. In 2015, a project arrangement on SFR System Integration and Assessment (SIA) has been signed by 7 members : China, EU, France, Japan, Korea, Russia and US. In the SIA project, RD needs from the SFR design will be shown to the RD project, and RD results from each RD project will be integrated into the designs.
Hayafune, Hiroki; Chikazawa, Yoshitaka; Kamide, Hideki; Iwasaki, Mikinori*; Shoji, Takashi*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 11 Pages, 2017/06
Design studies on a next generation sodium-cooled fast reactor (SFR) considering the safety design criteria (SDC) developed in the generation IV international forum (GIF) was summarized. To meet SDC including the lessons learned from the TEPCO's Fukushima Dai-ichi Nuclear Power Plants accident, the heat removal function was enhanced to avoid loss of the function even if any internal events exceeding design basis or severe external event happen. Several design options have been investigated and auxiliary core cooling system using air as ultimate heat sink has been selected as an additional cooling system regarding system reliability and diversification. Even though the next generation SFR already adopts seismic isolation system, main component designs have been improved considering revised earthquake conditions. For other external events, design measures for various external events are taken into account. Reactor building design has been improved and important safety components are diversified and located separately improving independency. Those design studies and evaluations on the next generation sodium-cooled reactor have contributed to the development of safety design guidelines (SDG) which is under discussion in the GIF framework.
Arai, Masanobu; Kunogi, Kosuke; Aizawa, Kosuke; Chikazawa, Yoshitaka; Takaya, Shigeru; Kubo, Shigenobu; Kotake, Shoji*; Ito, Takaya*; Yamaguchi, Akira*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04
Applications for maintenance program on piping support of prototype fast breeder reactor Monju are studied. Based on degradation mechanism, snubbers in Monju primary cooling system showed lifetime more than the plant lifetime of 30 years by experiments conservatively. For the first step during construction, visual inspection on accessible all supports could be available. In that visual inspection, mounting conditions and damages of all accessible supports could be monitored. One of major features of the Monju primary piping system is large thermal expansion due to large temperature difference between maintenance and operation conditions. Thanks to that large thermal expansion, integrity of the piping support could be monitored by measuring piping displacement. When technologies of piping displacement monitoring are matured in Monju, visual inspection on piping support could be shifted to piping displacement monitoring. At that stage, the visual inspection could be limited only on representative supports.
Chikazawa, Yoshitaka; Kubo, Shigenobu; Shimakawa, Yoshio*; Kaneko, Fumiaki*; Shoji, Takashi*; Nakata, Shuhei*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04
Sodium-cooled reactor (SFR) has superior characteristics thanks to sodium coolant features such as low pressure and high natural convection capability. Involving lessons learned from the 1F accident, requirements on design base DHRS have been modified. In that modification, safety requirements on design extended conditions have been clarified and sodium temperature criteria have been changed taking into account design margin even for design extended conditions. With the new DHRS configuration including ACS, designs of component cooling water system and emergency power supply have been updated.
Sasaki, Kota*; Yusa, Noritaka*; Wakai, Takashi; Hashizume, Hidetoshi*
Electromagnetic Nondestructive Evaluation (XVIII), p.244 - 251, 2015/06
Times Cited Count：0 Percentile：0.00(Mechanics)Nishimura, Akihiko; Takenaka, Yusuke*; Furuyama, Takehiro*; Shimomura, Takuya; Terada, Takaya; Daido, Hiroyuki
Journal of Laser Micro/Nanoengineering, 9(3), p.221 - 224, 2014/11
Times Cited Count：0 Percentile：0.00(Nanoscience & Nanotechnology)Heat resistant FBG sensors were developed by femtosecond laser processing to apply them to high temperature operated piping system of nuclear power plants. The FBG sensor was installed on the surface of a steel blade and a vibration test was conducted to detect the resonant vibration frequency of the vibrating blade. The FBG sensor had the heatproof performance at 600
C. A frequency stabilized sensing system using a tunable laser was tested for structural health monitoring in daily operation of nuclear power plants. The FBG sensor was installed on the surface of a steel blade for vibration induced strain measurements. Welding, brazing, soldering and noble metal powder adhesive were discussed for molding the FBG sensors.
Hagiwara, Hiroyuki; Kimura, Nobuyuki*; Onojima, Takamitsu; Nagasawa, Kazuyoshi*; Kamide, Hideki; Tanaka, Masaaki
JAEA-Research 2014-014, 178 Pages, 2014/09
JAEA-Research-2014-014.pdf:53.12MB
Thermal stratification in the upper plenum is one of the most important issues of a reactor vessel in sodium cooled fast reactor. The steep temperature gradient across the stratification interface may cause the thermal load against the reactor vessel wall. In this study, the water experiment was carried out using the 1/11 scale upper plenum model of the Japan sodium-cooled fast reactor (JSFR) in order to evaluate the thermal stratification under the natural circulation condition and a direct heat exchanger (DHX) operation condition. The temperature gradient under the natural circulation condition was approximately 1/3 times smaller than that under the forced circulation condition. In the DHX operation case, the steep temperature gradient occurred in the lower region of upper plenum due to the cold fluid from the outlet of DHX.
Mizuta, Shunji; ;
JNC TN9400 2000-048, 28 Pages, 2000/04
JNC-TN9400-2000-048.pdf:0.64MB
ODS (Oxide Dispersion Strengthened) ferritic-martainsitic steels are one of the most prospective cladding materials for advanced fast breeder reactors, since they are expected to have excellent swelling resistance and superior high temperature strength due to the finely distributed stable oxide particles(Y
O
). Properties and the tentative strength equations for ODS ferritic-martainsitic were proposed on the basis of the latest data to apply to the feasibility study of the sodium coolant MOX fuel plant. The items of equations are follows. (1)creep rupture strength (2)correction factor of creep rupture strength (in Na and in reactor) (3)outer surface eorrosion (Na) (4)inner surface corrosion (in MOX fuel pin) (5)thermal conductivity
Sakai, Takaaki; ; Ohshima, Hiroyuki; Yamaguchi, Akira
JNC TN9400 2000-033, 94 Pages, 2000/04
JNC-TN9400-2000-033.pdf:4.36MB
The feasibility study on several concepts for the commercial fast breeder reactor(FBR) in future has been conducted in JNC for the kinds of possible coolants and fuel types to confirm the direction of the FBR developments in Japan. ln this report, Lead and Lead-Bismuth eutectic coolants were estimated for the decay heat removal characteristics by the comparison with sodium coolant that has excellent features for the heat transfer and heat transport performance. Heavy liquid metal coolants, such as Lead and Lead-Bismuth, have desirable chemical inertness for water and atmosphere. Therefore, there are many economical plant proposals without an intermediate heat transport system that prevents the direct effect on a reactor core by the chemical reaction between water and the liquid metal coolant at the hypocritical tube fairer accidents in a steam generator. ln this study, transient analyses on the thermal-hydraulics have been performed for the decay heat removal events in "Equivalent plant" with the Lead, Lead-Bismuth and Sodium coolant by using Super-COPD code. And a resulted optimized lead cooled plant in feasibility study was also analyzed for the comparison. ln conclusion, it is become clear that the natural circulation performance, that has an important roll in passive safety characteristic of the reactor, is more excellent in heavy liquid metals than sodium coolant during the decay heat removal transients. However, we need to conform the heat transfer reduction by the oxidize film or the corrosion products expected to appear on the heat transfer surface in the Lead and Lead-Bismuth circumstance.
Kato, Atsushi; Ichikawa, Kenta*; Nakata, Shuhei*; Chikazawa, Yoshitaka; Ando, Masato
no journal, ,
In the conceptual study of a tank-type sodium-cooled fast reactor, a decay heat removal system concept was developed, taking into account the use of natural circulation heat removal and the diversity of heat removal methods, and a basic plant operation way after the reactor trip was established.
Nishimura, Akihiko; Shimomura, Takuya; Takenaka, Yusuke*; Terada, Takaya; Daido, Hiroyuki
no journal, ,
no abstracts in English
C pellet immersed in molten stainless steelTakai, Toshihide; Emura, Yuki; Yamano, Hidemasa
no journal, ,
Relocation behavior of BC-SS eutectic melt may affect neutronic reactivity of the degraded core. Corrosion behavior of BC pellet immersed in molten SS were investigated and its reaction rate constants were calculated as basic data to improve CDA analysis code. In this report, post-test analysis results of reaction zone and reduction amount of BC pellet were described.
Kato, Atsushi; Onoda, Yuichi; Miyagawa, Takayuki*; Endo, Junji*; Kubo, Koji*
no journal, ,
JAEA is studying 600 MWe pool-type sodium-cooled fast reactor. This report presents thermal hydraulic study in a reactor vessel and structural intactness evaluation in case of station black out and reactor trip.
