Yamaguchi, Tetsuji; Ohira, Saki; Hemmi, Ko; Barr, L.; Shimada, Asako; Maeda, Toshikatsu; Iida, Yoshihisa
Radiochimica Acta, 108(11), p.873 - 877, 2020/11
Genshiryoku Bakkuendo Kenkyu (CD-ROM), 27(1), p.34 - 36, 2020/06
no abstracts in English
Ueta, Shohei; Mizuta, Naoki; Fukaya, Yuji; Goto, Minoru; Tachibana, Yukio; Honda, Masaki*; Saiki, Yohei*; Takahashi, Masashi*; Ohira, Koichi*; Nakano, Masaaki*; et al.
Nuclear Engineering and Design, 357, p.110419_1 - 110419_10, 2020/02
The concept of a plutonium (Pu) burner HTGR is proposed to incarnate highly-effective Pu utilization by its inherent safety features. The security and safety fuel (3S-TRISO fuel) employs the coated fuel particle with a fuel kernel made of plutonium dioxide (PuO) and yttria stabilized zirconia (YSZ) as an inert matrix. This paper presents feasibility study of Pu burner HTGR and R&D on the 3S-TRISO fuel.
Aihara, Jun; Yasuda, Atsushi*; Ueta, Shohei; Ogawa, Hiroaki; Honda, Masaki*; Ohira, Koichi*; Tachibana, Yukio
Nihon Genshiryoku Gakkai Wabun Rombunshi, 18(4), p.237 - 245, 2019/12
Development of fabrication and inspection technologies of oxidation resistant fuel element for improvement of safety of high temperature gas-cooled reactors (HTGRs) in severe oxidation accident was carried out. Simulated coated fuel particles (CFPs), alumina particles, were over-coated with mixed powder of Si, C and small amount of resin to form over-coated particles, and over-coated particles were molded and hot-pressed to sinter simulated oxidation resistant fuel elements with SiC/C mixed matrix. Simulated oxidation resistant fuel elements with matrix whose Si/C mole ratio is 1.00 were fabricated. Failure fraction of CFPs in fuel elements is one of very important inspection subjects of HTGR fuel. It is essential that CFPs are extracted from fuel elements without additional failure. Development of method for extraction of CFPs was carried out. Desolation of SiC by KOH method or pressurized acidolysis method should be applied to extraction of CFPs.
Aihara, Jun; Honda, Masaki*; Ueta, Shohei; Ogawa, Hiroaki; Ohira, Koichi*; Tachibana, Yukio
Nihon Genshiryoku Gakkai Wabun Rombunshi, 18(1), p.29 - 36, 2019/03
Japan Atomic Energy Agency carried out development of fabrication technology of oxidation resistant fuel element for improvement of safety of high temperature gas-cooled reactors in serious oxidation accident, based on precursor research in former JAEA. Dummy coated fuel particles (alumina particles) were over-coated with mixed powder of Si, C and small amount of resin to form over-coated particles, and over-coated particles were molded and hot-pressed to sinter dummy oxidation resistant fuel elements with SiC/C mixed matrix. We fabricated dummy oxidation resistant fuel elements with matrix whose Si/C mole ratio (about 0.551) is three times as large as that in precursor research. Si peak was not detected by X-ray diffraction of matrix. Better oxidation resistant was confirmed with oxidation test in 20% O at 1673 K than that of ordinal fuel compact with ordinal graphite/carbon matrix. All dummy coated fuel particles were held in specimen after 10 h oxidation.
Nabeshima, Kunihiko; Doda, Norihiro; Ohshima, Hiroyuki; Mori, Takero; Ohira, Hiroaki; Iwasaki, Takashi*
Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.1041 - 1049, 2015/08
Natural circulation is one of the most important mechanisms to remove decay heat in the sodium cooled fast reactors from the viewpoint of passive safety. On the other hand, it is difficult to evaluate plant dynamics accurately under low flow natural circulation condition. In this study, Super-COPD has been validated through the application to the analysis of natural circulation tests in the experimental fast reactor JOYO. Almost all plant components in JOYO including four air-coolers were modeled in Super COPD. Furthermore, the full scale modeling of fuel subassembly was also adopted in this analysis. The natural circulation test after reactor scram from 100 MW full power at JOYO was selected and simulated by Super-COPD. The transient behaviors predicted by Super-COPD showed good agreement with the experimental data.
Ohira, Hiroaki; Doda, Norihiro; Kamide, Hideki; Iwasaki, Takashi*; Minami, Masaki*
Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.2585 - 2592, 2015/05
IAEA's Coordinated Research Project on Benchmark Analyses of Shutdown Heat Removal Test (SHRT) performed at the Experimental Breeder Reactor-II (EBR-II) has been carried out since 2012. The benchmark specifications were provided by the Argonne National Laboratory (ANL) and the model development for thermal-hydraulics codes and/or plant dynamics codes has been conducted by participating organizations. The experimental data were also provided by the ANL after the calculations have been performed as the blind simulation. JAEA participated in this benchmark analyses, and the plant dynamics analysis code; Super-COPD was applied to the SHRT-17 simulation. The calculated inlet temperature of the high pressure plenum agreed well with the test data in all simulation time. Although the Z-pipe inlet temperature and the IHX intermediate outlet temperature had some discrepancy in the first 400 sec. caused by larger mass flow rate of the primary pump and the perfect mixing model of upper plenum, these temperatures and the flow rate agreed well with the measured data after 400 sec. Hence it was concluded the present analytical model could predict the natural circulation in good accuracy.
Doda, Norihiro; Igawa, Kenichi*; Minami, Masaki*; Iwasaki, Takashi*; Ohira, Hiroaki
no journal, ,
Sodium-cooled fast reactors have been developed aiming at introducing natural circulation decay heat removal systems by utilizing the characteristic of having a large coolant temperature difference between at the inlet and at the outlet of reactor vessel. In this study, as part of validation for core hot spot evaluation method, which is required for adoption of natural circulation decay heat removal systems, EBR-II (Experimental Breeder Reactor II) shutdown heat removal test was simulated. The simulation results demonstrated that the evaluation method sufficiently predicts the whole plant thermal hydraulic behaviors and the maximum coolant temperature in a fuel subassembly in natural circulation decay heat removal.
Honda, Masaki*; Yasuda, Atsushi*; Ohira, Koichi*; Tachibana, Yukio
no journal, ,
For the purpose of upgrading safety of High Temperature Gas-cooled Reactor (HTGR), research on inspection technology for advanced fuel element was conducted with cooperation of JAEA from FY2014 to 2016. The advanced fuel element contains SiC as the matrix material and oxidation resistance is highly improved so that shape and integrity of the fuel element should be maintained even when large unexpected amount of air enters into the core in the air ingress (pipe rupture) accident which is a typical event for the HTGR. For the inspection technology, evaluation and inspection methods on homogeneity of SiC/C matrix and distribution of coated fuel particles as well as dissolution methods of coated fuel particles and SiC/C matrix for evaluation of failure fraction were developed and established.
Yamaguchi, Tetsuji; Hemmi, Ko; Logan, B.*; Shimada, Asako; Ohira, Saki; Iida, Yoshihisa
no journal, ,
We developed a sorption model taking into account the surface species of Ca-Nb-OH, and reanalyzed the data of Ervanne et al. (2014) for illite. As a result, we were able to reproduce the high sorption distribution coefficient of Nb to illite in the presence of Ca.
Shimada, Asako; Taniguchi, Yoshinori; Ohira, Saki; Iida, Yoshihisa
no journal, ,
no abstracts in English