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Yamashita, Shinichiro; Nagase, Fumihisa; Kurata, Masaki; Kaji, Yoshiyuki
Proceedings of Annual Topical Meeting on LWR Fuels with Enhanced Safety and Performance (TopFuel 2016) (USB Flash Drive), p.21 - 30, 2016/09
Fuel rod, channel box, and control rod designed with new materials and concepts have been developed in Japan for increasing accident tolerance of LWRs. In order to efficiently and properly implement the accident tolerant fuels (ATFs) and the other components, it is necessary not only to accumulate fundamental and practical data but also to consider technology readiness, recognize knowledge gaps, and establish strategy for design and fabrication. The Japan Atomic Energy Agency (JAEA) has established the above "technical basis" and drafted a research plan towards implementation of the ATFs and components as a program sponsored and organized by the Ministry of Economy, Trade and Industry (METI). It is useful to take advantage of the experiences in commercial uses of zirconium-base alloys in LWRs and, therefore, JAEA has conducted this METI project in cooperation with power plant providers, fuel venders, research institutes and universities who have been involved in the development of the ATF materials. The present paper describes the main results of the project conducted to establish the technical basis of the ATFs and components.
Nagase, Fumihisa; Ishikawa, Jun; Kurata, Masaki; Yoshida, Hiroyuki; Kaji, Yoshiyuki; Shibamoto, Yasuteru; Amaya, Masaki; Okumura, Keisuke; Katsuyama, Jinya
Proceedings of International Nuclear Fuel Cycle Conference; Nuclear Energy at a Crossroads (GLOBAL 2013) (CD-ROM), p.711 - 720, 2013/09
Estimation of the accident progress and status inside the reactor is required to properly and reliably conduct decommissioning of the Fukushima-Daiichi NPPs. For that, it is necessary to obtain additional experimental data and revised models for the estimation using computer codes with increased accuracies. JAEA has selected phenomena to be reviewed and developed in terms of thermo hydraulic behavior in the reactor, progression of fuel bundle degradation, failure of the lower head of the reactor pressure vessel, and analysis of the accident, considering previously obtained information, conditions specific to the Fukushima-Daiichi NPP accident, and recent progress of experimental and analytical technologies. This paper introduces the selected phenomena to be reviewed and developed and recent results from the JAEA's corresponding research programs.
Watanabe, Takeshi*; Tsushima, Satoru*; Yamamoto, Ichiro*; Tomioka, Osamu; Meguro, Yoshihiro; Nakashima, Mikio; Wada, Ryutaro*; Nagase, Yoshiyuki*; Fukuzato, Ryuichi*
Proceedings of 2nd International Symposium on Supercritical Fluid Technology for Energy and Environment Applications (Super Green 2003), p.363 - 366, 2004/00
Recovery of salts by supercritical fluid leaching (SFL) method using carbon dioxide was experimentally studied. It was confirmed that LiCl was recovered with a mixed fluid of carbon dioxide and methanol, and KCl and SrCl
were recovered with a mixed fluid of carbon dioxide, methanol and crown ether. The influence of crown ether for KCl and SrCl extraction were found to increase in the order of 15-crown-5 (15C5) 
18-crown-6 (18C6) dicychlohexyl-18-crown-6 (DC18C6). It is expected that other salts can be recovered selectively with a mixed fluid of carbon dioxide, methanol and suitable crown ether.
Nagase, Yoshiyuki*; Masuda, Kaoru*; Wada, Ryutaro*; Yamamoto, Ichiro*; Tomioka, Osamu; Meguro, Yoshihiro; Fukuzato, Ryuichi*
Proceedings of 2nd International Symposium on Supercritical Fluid Technology for Energy and Environment Applications (Super Green 2003), p.254 - 257, 2004/00
no abstracts in English
Kaji, Yoshiyuki; Katsuyama, Jinya; Yamaguchi, Yoshihito; Abe, Yosuke; Nemoto, Yoshiyuki; Nagase, Fumihisa
no journal, ,
The detailed 3D model of reactor pressure vessel lower head with control rod guide tubes(CRGTs) and shroud supports are constructed with considering weld portions of penetrations of CRGTs as based on the open-public literatures. This paper describes the thermal hydraulic and structural analysis results for failure behaviors of BWR pressure vessel lower head.
Nemoto, Yoshiyuki; Kato, Hitoshi; Kaji, Yoshiyuki; Nagase, Fumihisa
no journal, ,
Using finite element method (FEM) analysis, JAEA studies on failure of reactor pressure vessel lower head due to the sever accidents in Fukushima. Multi-axial stress condition should be taken account in evaluation of material deformation in complicated structure, e.g., joint parts of lower head and control rod guide tube, however material property data obtained by former uniaxial material testing is applied in the FEM analysis. In this study, internal pressure creep test and its FEM modeling analysis were conducted to evaluate applicability of uniaxial data for analysis of creep behavior under multi-axial stress condition.
Katsuyama, Jinya; Yamaguchi, Yoshihito; Kaji, Yoshiyuki; Nagase, Fumihisa
no journal, ,
no abstracts in English
Kaji, Yoshiyuki; Katsuyama, Jinya; Yamaguchi, Yoshihito; Nemoto, Yoshiyuki; Abe, Yosuke; Nagase, Fumihisa
no journal, ,
In order to predict the accident progress and inside reactor situation in the Fukushima Dai-ichi Nuclear Power Plant, we perform the research activities about failure evaluation of reactor pressure vessel lower head by molten fuel. In this presentation, we describe high temperature materials data and detailed analytical model for lower head, and introduce examples of analytical results of coupled between thermal-hydraulic and structural analysis for reactor pressure vessel lower head due to severe accident.
Kaji, Yoshiyuki; Katsuyama, Jinya; Yamaguchi, Yoshihito; Nemoto, Yoshiyuki; Abe, Yosuke; Nagase, Fumihisa
no journal, ,
In existing severe accident code, rupture of reactor pressure vessel (RPV) lower head after melt down of core is analyzed using the simple model like the Larson-Miller model. It is difficult to evaluate the local deformation and rupture behavior for the actual lower head with such a simple model. Therefore, in order to predict the real position of molten fuel outside pressure vessel, it is necessary to evaluate rupture time and rupture behavior of RPV lower head of BWR precisely.Re-evaluation of materials data such as mechanical properties, creep deformation/rupture properties is made for low alloy steel, Ni-based alloy and stainless steels based on past research activities. To expand materials database and verify the creep constitutive equation and rupture model, we started obtaining the materials data under uniaxial and multi-axial stress conditions at high temperature near melting point. To investigate the inhomogeneous temperature and stress distribution by geometrical complex of BWR lower head, the detailed 3D model of RPV lower head with control rod guide tubes (CRGTs) and shroud supports are constructed and the 3D thermal hydraulic analysis of simulated molten pool and creep deformation analysis of lower head are performed using ANSYS Fluent / Mechanical finite element code. It is found that the possibility of failure mode for BWR lower head are both the penetration failure which is melt-through or drop-away of the guide tube, local rupture and global rupture of lower head by creep deformation mechanism and the melting collapse mechanism due to different boundary conditions.
Nagase, Fumihisa; Yoshida, Hiroyuki; Kaji, Yoshiyuki; Amaya, Masaki
no journal, ,
In order to conduct decommissioning of the Fukushima-Daiichi NPPs properly and reliably, increase of accuracy is required in estimation of the accident progress and status inside the reactors. It is also required to confirm effectiveness of safety features against prevention and mitigation of severe accident (AM) in the existing power reactors. Furthermore, it is necessary to prepare own prediction methods to demonstrate prototypical accident phenomena with higher accuracy. Therefore, Japan Atomic Energy Agency (JAEA) conducts researches and developments in terms of thermal hydraulic behaviour in reactor, fuel damage and degradation process, behaviour of structural materials and pressure vessel, accident analysis, and molten materials relocation in the lower plenum region between lower core plate and lower head in BWRs during severe accidents. The paper presents progresses and recent results of the main research subjects.
Nagase, Fumihisa; Yoshida, Hiroyuki; Nemoto, Yoshiyuki; Amaya, Masaki; Yamashita, Shinichiro
no journal, ,
In order to conduct decommissioning of the Fukushima-Daiichi NPP properly and reliably, increase of accuracy is required in estimation of the accident progress and status inside the reactors. It is also required to improve analytical methods to estimate severe accident progression in the existing light water reactors. Therefore, JAEA conducts fundamental studies in terms of thermal hydraulic behaviour in reactor, fuel damage and degradation process, behaviour of structural materials and pressure vessel, release and migration of fission products, and so on under severe accident conditions. The present paper reports overview of the fundamental studies and recent results.
Nagase, Fumihisa; Yoshida, Hiroyuki; Nemoto, Yoshiyuki; Amaya, Masaki; Yamashita, Shinichiro
no journal, ,
In order to conduct decommissioning of the Fukushima-Daiichi NPP properly and reliably, increase of accuracy is required in estimation of the accident progress and status inside the reactors. It is also required to improve analytical methods to estimate severe accident progression in the existing light water reactors. Therefore, JAEA conducts fundamental studies in terms of thermal hydraulic behaviour in reactor, fuel damage and degradation process, behaviour of structural materials and pressure vessel, release and migration of fission products, and so on under severe accident conditions. The present paper reports overview of the fundamental studies and recent results.
Yamashita, Shinichiro; Nagase, Fumihisa; Kurata, Masaki; Kaji, Yoshiyuki
no journal, ,
In TEPCO's Fukushima Daiichi Nuclear Power Station (1F), the cooling capability was lost due to tsunami caused by the Great East Japan Earthquake. It is considered that the Zirconium (Zr) alloy fuel cladding was oxidized in the heated core and the subsequent temperature escalation due to the oxidation caused core melting. As lessons learned from the 1F accident, development of advanced fuel and core components with enhanced accident tolerance becomes the greater concern. We have started R&D to establish the technical basis to implement the advanced fuel components to existing LWRs. The R&D is conducted in cooperation with power plant providers, fuel venders, research institutes and universities who have been involved in the development of the advanced fuel components. In this presentation, we will introduce the R&D program which are carried out under the Project on Development of Technical Basis for Safety Improvement at Nuclear Power Plants by Ministry of Economy, Trade and Industry (METI) of Japan.