Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Onitsuka, Takashi; Nogiwa, Kimihiro; Abe, Teruyoshi; Sakakibara, Yasuhide; Nakamura, Takahisa; Horie, Kaoru
no journal, ,
no abstracts in English
Sawaguchi, Takuma
no journal, ,
no abstracts in English
Ishigaki, Masahiro
no journal, ,
no abstracts in English
Sugiyama, Tomoyuki; Fukuda, Takuji; Nagase, Fumihisa
no journal, ,
To confirm the reactor safety design, safety reviews are performed under accident conditions as well as under normal operation conditions. One of such design basis accidents is the reactivity-initiated accident (RIA); a reactor excursion typically caused by rapid ejection of the control rod. In case of high burnup fuel, mechanical failure of corroded cladding can occur due to the rapid thermal expansion of fuel pellet. Cladding mechanical test is effective to evaluate the failure limit, but the conventional tests using axial tension or hydraulic pressure cannot represent the biaxial stress state during RIA, which is generated in the cladding chemically bonded with the fuel pellet. Therefore, the biaxial stress testing apparatus was developed, which simulates the stress condition in RIA by the simultaneous axial load and hydraulic pressurization. Mechanical properties of zircaloy-4 cladding are being acquired with the apparatus to improve the accuracy of fuel failure limit evaluation.
Shibamoto, Yasuteru
no journal, ,
no abstracts in English
Takeda, Seiji
no journal, ,
no abstracts in English
Suyama, Kenya
no journal, ,
In order to estimate the characteristic (criticality, activity, decay heat etc.) of the spent nuclear fuel, it is necessary to evaluate its isotopic composition accurately. For that purpose, our research group has been working on the development of SWAT code system for the accurate and precise computer simulation, obtaining the experimental data of assay data of spent nuclear fuel by the destructive experiment, and compiling the database for sharing the isotopic composition of the spent nuclear fuel in the international community.
Kimura, Masanori
no journal, ,
no abstracts in English
Chimi, Yasuhiro; Nishiyama, Yutaka
no journal, ,
In the Fuels and Materials Irradiation Research Group, in order to provide the technical information for the safety regulation of the Japanese government on the ageing management technical evaluation for the current commercial light water reactors (LWRs), irradiation studies on the structural materials for LWRs (i.e. irradiation embrittlement of reactor pressure vessels and irradiation-assisted stress corrosion cracking (IASCC) of reactor core shroud) under simulated LWR water and irradiation conditions will be performed by using the Japan Materials Testing Reactor (JMTR), which will restart in FY 2012.
Katsuyama, Jinya; Yamaguchi, Yoshihito; Masaki, Koichi; Onizawa, Kunio
no journal, ,
It is important to accurately assess the structural integrity considering the material aging degradation of pressure boundary piping and reactor pressure vessel (RPV) in nuclear power plants. For this purpose we have studied on the advanced methodologies for the structural integrities of piping and RPV. The effect of excessive loading to the piping such as a large earthquake on the residual stress has been evaluated by the finite element method. A crack growth evaluation method applicable to the excessive loading has been proposed through laboratory experiments. A probabilistic fracture mechanics (PFM) analysis code for piping was improved to incorporate the proposed method. A fracture mechanics analysis method considering the inhomogeneity of irradiation susceptibility and microstructure of heat affected zone in RPV has also been developed. For the rationalization of the structural integrity assessment of aged RPVs, a PFM analysis code for RPV was improved to incorporate the method.
Watanabe, Tadashi
no journal, ,
The analysis of long-term station blackout accident of BWR has been performed using TRAC-BF1 code. The actuation of RCIC was assumed, and the results were compared with the observed data at the Fukushima Daiichi power plant unit 2 reactor. The effectiveness of recovery action for reactor cooling was discussed after the termination of RCIC. A BWR-5 with 1100 MW was analyzed, while the unit 2 was a BWR-4 with 780 MW. The reactor pressure and the core liquid level were, however, in good agreement with the observed data. It was confirmed that the quasi-steady state was continued for a long time by the RCIC actuation. The timing of recovery action, which is composed of depressurization and coolant injection, necessary for the clad temperature being less than 1500 K was studied and compared with the unit 2. The effect of containment on the core thermal hydraulics is also discussed.
Ishikawa, Jun
no journal, ,
no abstracts in English
Fuketa, Toyoshi
no journal, ,
no abstracts in English
