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Hirooka, Shun; Kato, Masato; Watanabe, Masashi
Transactions of the American Nuclear Society, 118, p.1624 - 1626, 2018/06
This study suggested the time development of oxygen-to-metal ratio (O/M) redistribution model with oxygen-related properties in MOX. Irradiation simulation including the suggested O/M redistribution and pore migration with vaporization-condensation model which bares density redistribution was demonstrated. The simulation results showed that O/M redistribution proceeded at lower temperature than density redistribution, which indicated that oxygen diffusion got influential at lower temperature than vaporization-condensation of MOX. Another find was that O/M redistribution was very slow at the surface because temperature kept low. However, near the surface (inside from the surface) where the temperature exceeded 1000 K, O/M redistribution was rather recognizable with oxygen flown from inner region to the near-surface. The results will be evaluated by comparison with post-irradiation examination data.
pellets by 28 GHz microwave irradiationAkashi, Masatoshi; Matsumoto, Taku; Kato, Masato
Transactions of the American Nuclear Society, 118, p.1391 - 1394, 2018/06
In this study, CeO pellet sintering by irradiating microwave at a frequency of 28 GHz was carried out to investigate the effect of particle diameter of raw powder on the density of sintered pellet. The highest bulk density is 94.2 %T.D. under the condition of 30 min holding at 1473 K. The bulk density decreases with increasing the particle diameter of used raw powder. On the other hand, all of the apparent density of sintered pellet is more than 93.5 %T.D.. The difference between the bulk density and the apparent density is caused by the difference of open porosity for each sample pellet. It seems that the high density sintered pellets with porous structure are obtained because sample pellet is heated internally and uniformly in microwave sintering.
Kato, Masato
Transactions of the American Nuclear Society, 114, p.987 - 988, 2016/06
In Japan, uranium and plutonium mixed oxide (MOX) has been developed as fuels of sodium-cooled fast reactors. The developing MOX fuels come in variety of O/M ratio, Pu content, minor actinide (MA) content and density. We have studied a science based fuel technology to evaluate fuel behaviors in fabrication process and irradiation condition of such various fuels. The technologies which are constructed based on experimental database can apply to mechanistic evaluation of fuel behaviors. To develop the science based fuel technology, many different varieties of basic properties have been investigated, and experimental database was constructed. And a mechanistic physical property model has been studied. The models contribute to describe various behaviors in fuel fabrication process and irradiation condition.
Watanabe, Masashi; Kato, Masato; Sunaoshi, Takeo*
Transactions of the American Nuclear Society, 114, p.1081 - 1082, 2016/06
Many studies on the oxygen potential of UO have been carried out so far. However, the oxygen potential data for UO near the stoichiometric composition in the high temperature region (1673-1873 K) are limited. In this work, the oxygen potential data of UO
were extended to high temperature range of 1673-1873 K by gas equilibrium method. The measured data were analyzed based on a defect chemistry model.
4 bundle at 2 MPaLiu, W.; Nagatake, Taku; Shibata, Mitsuhiko; Takase, Kazuyuki; Yoshida, Hiroyuki
Transactions of the American Nuclear Society, 114, p.875 - 878, 2016/06
To contribute to the clarification of the Fukushima Daiichi Accident, JAEA is working on getting instantaneous void fraction distribution data in steam water two - phase flow in rod bundle geometry under high pressure, high temperature condition, with using Wire Mesh Sensor (WMS) developed at JAEA for high pressure, high temperature condition, focusing on the low flow rate condition after the reactor scram. This paper reports the experimental results for the measured void fraction distribution in steam vapor two-phase flow in a 4 4 bundle under 1.6 MPa (202 
C), 2.1 MPa (215 C) and 2.6 MPa (226 C) conditions. The data is expected to be used in the validation of the detailed two-phase flow codes TPFIT and ACE3D developed at JAEA. The time and space averaged void fraction data is also expected being used in the validation of the drift flux models implemented in the two fluids codes, such as TRACE code.
Aoyama, Takafumi; Ishikawa, Takashi; Ito, Chikara; Iwata, Yoshihiro; Harano, Hideki*
Transactions of the American Nuclear Society, 106(1), p.611 - 613, 2012/06
We studied the failed fuel detection and location (FFDL) system by means of laser resonance ionization mass spectrometry (RIMS). The applicability of the RIMS system for Monju and its installation were investigated in this study. It was found that the RIMS is able to identify the neighboring tag gas which is used for Monju with the 68% reliability. The newly designed RIMS system can be accommodated in the present facility and can be connected to primary cover gas lines without affects to the current FFDL function of the conventional system.
Furukawa, Tomohiro; Kato, Shoichi; Yoshida, Eiichi
Journal of Nuclear Materials, 392(2), p.249 - 254, 2009/07
Times Cited Count：41 Percentile：92.56(Materials Science, Multidisciplinary)In order to incorporate an evaluation procedure for sodium environmental effects of core and structural materials into the elevated temperature structural design guide for fast breeder reactors (FBRs), research and development of sodium compatibility of the materials has been performing in Japan Atomic Energy Agency. This paper describes the overview of sodium compatibility of the candidate materials which are used as core and structures of Japanese demonstration FBRs.
Umeda, Miki; Ueta, Shohei; Sugiyama, Tomoyuki
Transactions of the American Nuclear Society, 98(1), P. 987, 2008/06
Konomura, Mamoru; Ichimiya, Masakazu
Journal of Nuclear Materials, 371(1-3), p.250 - 269, 2007/09
Times Cited Count：18 Percentile：75.3(Materials Science, Multidisciplinary)no abstracts in English
Katsuyama, Kozo; Nagamine, Tsuyoshi; Nakamura, Yasuo; Matsumoto, Shinichiro; Asaka, Takeo; Furuya, Hirotaka
Transactions of the American Nuclear Society, 94(1), p.771 - 772, 2006/06
no abstracts in English
Sawa, Kazuhiro; Ueta, Shohei; Aihara, Jun
Transactions of the American Nuclear Society, 94(1), P. 705, 2006/06
The Very-High-Temperature Reactor (VHTR) is the one of the most promising candidates for the Generation IV Nuclear Energy System. The VHTR fuel should exhibit excellent safety performance up to burn-up of about 15 to 20 % fissions per initial metal atom (FIMA) and fluence of 610
m
(E
0.1MeV). There is no experimental data which has proved the intactness of conventional SiC-coated fuel particles under such severe condition. Japan Atomic Energy Agency (JAEA) has developed Zirconium carbide (ZrC)-coated fuel particle, the ZrC coating layer of which is expected to maintain its intactness under higher temperature and burn-up compared with SiC-coating layer. JAEA started (1) ZrC-coating process development by large-scale coater, (2) inspection method development of ZrC coating and (3) irradiation test and post irradiation experiment of ZrC coated particles. This paper presents present status of ZrC-coated fuel particle development in JAEA.
Nakamura, Kinya*; Yokoo, Takeshi*; Arai, Yasuo
Transactions of the American Nuclear Society, 94(1), P. 780, 2006/06
no abstracts in English
Murakami, Takashi; Kimura, Hideo
Materials Research Society Symposium Proceedings, Vol.294, p.535 - 542, 1993/00
no abstracts in English
Onuki, Toshihiko; Murakami, Takashi; Yanase, Nobuyuki
Materials Research Society Symposium Proceedings, Vol.294, p.527 - 533, 1993/00
no abstracts in English
Ukai, Shigeharu; Kaito, Takeji; Otsuka, Satoshi; Narita, Takeshi; Sakasegawa, Hideo
no journal, ,
The composition of martensitic 9Cr-ODS steels was optimized from the balancing of the residual alpha hard and F/M soft grains. The claddings with the optimized composition of 9Cr-0.14C-2W-0.3Ti-0.35YO
as well as standard were manufactured and their mechanical properties satisfied the design criteria for SFR fuel.
Nabielek, H.*; Verfondern, K.*; Tang, C.*; Ueta, Shohei
no journal, ,
Burn-leach is the most sensitive method for the determination of High-Temperature Gas-Cooled Reactor fuel quality. German fuel manufacture was for the operation of the 46 MWth AVR. Improvements in the fuel quality were due to perfected tabling of kernels, particles and overcoated particles and the introduction of automated overcoating. Chinese HTGR first load fuel manufacture around 2000 was for the operation of the 10 MWth HTR-10. An improvement can be observed after the first few production runs. Japanese HTGR first fuel manufacture was for the operation of the 30 MWth HTTR. The particle volume density of 30 % is much higher than the below 10 % of the spherical fuel elements. Nevertheless, very good results in terms of low defect fractions were also achieved. These results establish the quality standard in modern UO Triso fuel.
Morishita, Masaki; Asayama, Tai; Inoue, Masaki; Kotake, Shoji; Mizuno, Tomoyasu
no journal, ,
JSFR is a sodium cooled loop type fast reactor on which a conceptual design study is now underway in the framework of "Fast Reactor Cycle Technology Development Project (FaCT Project)" of Japan. Achieving economic competitiveness with future light water reactors, along with assuring high level of safety and reliability, is among its most crucial development targets. In this paper described are the major design features of JSFR and related material development issues.
Funasaka, Hideyuki
no journal, ,
The following 3 subjects is introduced and discussed in this panel session of ANS Meeting. (1) Current status and perspective of Fast Reactor Cycle Technology Development Project. (2) Related facilities for closed fuel cycle demonstaration and implementation. (3) Existing collaborative agreements for fuel cycle development and potential for new collaboration in future.
Hinoki, Tatsuya*; Park, Y.*; Park, J.*; Miwa, Shuhei; Donomae, Takako
no journal, ,
Nitride fuel and oxide fuel are candidate for helium gas-cooled fast reactor to achieve thermal efficiency over 40%. The conceptual design of fuel, which consists of nitride fuel of around 50%, highly dense SiC matrix and porous SiC buffer layers is encouraged according to the eport for "Feasibility Study on Commercialized Fast Reactor Cycle System (FS)", Japan. However there is no technique to form dense SiC in the small gap between fuels with accurate arrangement of fuels. The objective of this work is to develop basic fabrication technique for the high burn-up fuel utilizing forming technique for highly dense NITE-SiC/SiC composites. The cylindrical fuel is proposed in this work instead of spherical fuel to achieve high dense fuel and accurate arrangement.
Furukawa, Tomohiro; Kato, Shoichi; Yamamoto, Masaya
no journal, ,
Zirconium alloy has better neutron economy in most of the energy region of the fast reactor core. Therefore, the installation of zirconium reflectors to the Joyo has been planned to improve the neutron efficiency and increase the core average burn-up. As one step for achieving the plan, it is necessary to estimate the corrosion resistance of the alloy in the coolant sodium. So, authors have been firstly performed the compatibility test in stagnant sodium, and obtained that the alloy shows good corrosion resistance in high temperature. However, since the experiment was carried out in stagnant condition, thermal gradient mass transfer which was a main corrosion factor in liquid metals could not be taken into consideration. In this study, compatibility test of the alloy in a sodium test loop which has thermal gradient effect has been performed as the 2nd experiment. After the testing, metallurgical examination and mechanical strength testing were performed for the immersed specimens.
