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CPham, V. H.; Kurata, Masaki; Nagae, Yuji; Ishibashi, Ryo*; Sasaki, Masana*
Corrosion Science, 255, p.113098_1 - 113098_9, 2025/10
Times Cited Count：1 Percentile：0.00(Materials Science, Multidisciplinary)Wada, Yuki; Shibamoto, Yasuteru; Hibiki, Takashi*
International Journal of Heat and Mass Transfer, 249, p.127219_1 - 127219_16, 2025/10
Times Cited Count：0 Percentile：0.00(Thermodynamics)Sato, Rika; Kondo, Toshiki; Umeda, Ryota; Kikuchi, Shin; Yamano, Hidemasa
Progress in Nuclear Science and Technology (Internet), 8, p.137 - 142, 2025/09
In a sodium-cooled fast reactor (SFR) coupled to thermal energy storage (TES) system, the reaction between nitrate molten salt as thermal energy storage medium and sodium (Na) as reactor coolant might occur under postulated accidental conditions. Thus, the reaction behavior of Na-nitrate molten salt is one of the important phenomena in terms of safety assessment of the SFR with TES system. In this study, reaction experiments on Na-solar salt were performed. It was found that Na-solar salt reaction occurred after the NaNO
-KNO eutectic melting. Based on the measured reaction temperature, the kinetic parameters and rate constant were obtained and compared with the sodium-water reaction. From the results of kinetic analysis, it could be assumed that Na-solar salt reaction occurs in the time frame of the accident such as the failure of heat transfer tube of sodium-molten salt heat exchanger.
Yoshida, Kazuo; Hiyama, Mina*; Tamaki, Hitoshi
JAEA-Research 2025-003, 24 Pages, 2025/06
JAEA-Research-2025-003.pdf:2.06MB
An accident of evaporation to dryness by boiling of high-level radioactive liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (RuO
) are released from the tanks with water and nitric-acid mixed vapor into the atmosphere. Accurate quantitative estimation of released Ru is one of the important issues for risk assessment of those facilities. RuO is expected to be absorbed chemically into water dissolving nitrous acid. Condensation of mixed vapor plays an important role for Ru transporting behavior in the facility building. The thermal-hydraulic behavior in the facility building is simulated with MELCOR code. The latent heat, which is a governing factor for vapor condensing behavior, has almost same value for nitric acid and water at the temperature range under 120 centigrade. Considering this thermal characteristic, it is assumed that the amount of nitric acid is substituted with mole-equivalent water in MELCOR simulation. Compensating modeling induced deviation by this assumption have been assembled with control function features of MELCOR. The comparison results have been described conducted between original simulation and modified simulation with compensating model in this report. It has been revealed that the total amount of pool water in the facility was as same as both simulations.
Wada, Yuki; Shibamoto, Yasuteru; Hibiki, Takashi*
International Journal of Heat and Mass Transfer, 239, p.126598_1 - 126598_18, 2025/04
Times Cited Count：4 Percentile：41.94(Thermodynamics)Sakurai, Junya*; Torigata, Keisuke*; Matsunaga, Manabu*; Takanashi, Naoto*; Hibino, Shinya*; Kizu, Kenichi*; Morita, Akira*; Inomoto, Masahiro*; Shimohata, Nobuaki*; Toyota, Kodai; et al.
Tetsu To Hagane, 111(5), p.246 - 262, 2025/04
Tashiro, Shinsuke; Uchiyama, Gunzo; Ono, Takuya; Amano, Yuki; Yoshida, Ryoichiro; Watanabe, Koji*; Abe, Hitoshi; Yamane, Yuichi
Nuclear Technology, 211(3), p.429 - 438, 2025/03
Times Cited Count：1 Percentile：37.73(Nuclear Science & Technology)Contributing to the confinement safety evaluation of glove-box (GB) connected with high efficiency particle air (HEPA) filters for radioactive materials under fire accidents, combustion tests of a flammable polymer, Polymethyl methacrylate (PMMA), and a flame retardant polymer, Polycarbonate (PC), as typical GB panel resins have been conducted with an engineering-scale combustion apparatus. The combustion properties such as the mass loss rate (MLR) and the heat release rate (HRR) of PMMA and PC were investigated in the combustion tests. In the tests with the same shapes, it was found the followings; MLRs and HRRs of PMMA were larger than those of PC under the same supply flow rate into the combustion cell (Fv); MLRs and HRRs of PMMA and PC were constant under different Fv. Moreover, in the tests of PMMA with different cross section areas (S), MLRs and HRRs were found to be proportional to S. Using these results, the relationships of MLR and HRR to S of PMMA and PC were deduced.
Aihara, Jun; Ueta, Shohei; Honda, Masaki*; Kasahara, Seiji; Okamoto, Koji*
JAEA-Research 2024-012, 98 Pages, 2025/02
JAEA-Research-2024-012.pdf:32.24MB
Concept of Pu-burner high temperature gas-cooled reactor (HTGR) was proposed for the purpose of more safely reducing amount of recovered Pu. In Pu-burner HTGR concept, coated fuel particle (CFP), with ZrC coated yttria stabilized zirconia (YSZ) containing PuO
(PuO-YSZ) small particle and with tri-structural isotropic (TRISO) coating, is employed for very high burn-up and high nuclear proliferation resistance. ZrC layer is oxygen getter. In research project of Pu-burner HTGR carried out from fiscal year of 2014 to fiscal year of 2017, simulated CFPs were fabricated using Ce to simulate Pu. Moreover, simulated fuel compacts were fabricated using fabricated simulated CFPs. In this report, results of microstructural observation of CeO-YSZ and ZrC layer at each fabrication step are reported.
Okita, Shoichiro; Aoki, Takeshi; Fukaya, Yuji; Tachibana, Yukio
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 5 Pages, 2024/11
Brear, D. J.*; Kondo, Satoru; Sogabe, Joji; Tobita, Yoshiharu*; Kamiyama, Kenji
JAEA-Research 2024-009, 134 Pages, 2024/10
JAEA-Research-2024-009.pdf:2.45MB
The SIMMER-III/SIMMER-IV computer codes are being used for liquid-metal fast reactor (LMFR) core disruptive accident (CDA) analysis. The sequence of events predicted in a CDA is often influenced by the heat exchanges between LMFR materials, which are controlled by heat transfer coefficients (HTCs) in the respective materials. The mass transfer processes of melting and freezing, and vaporization and condensation are also controlled by HTCs. The complexities in determining HTCs in a multi-component and multi-phase system are the number of HTCs to be defined at binary contact areas of a fluid with other fluids and structure surfaces, and the modes of heat transfer taking into account different flow topologies representing flow regimes with and without structure. As a result, dozens of HTCs are evaluated in each mesh cell for the heat and mass transfer calculations. This report describes the role of HTCs in SIMMER-III/SIMMER-IV, the heat transfer correlations implemented and the calculation of HTCs in all topologies in multi-component, multi-phase flows. A complete description of the physical basis of HTCs and available experimental correlations is contained in Appendices to this report. The major achievement of the code assessment program conducted in parallel with code development is summarized with respect to HTC modeling to demonstrate that the coding is reliable and that the model is applicable to various multi-phase problems with and without reactor materials.
Rochman, D.*; Minato, Futoshi; Watanabe, Tomoaki; 52 of others*
EPJ Nuclear Sciences & Technologies (Internet), 10, p.9_1 - 9_83, 2024/10
Motegi, Kosuke; Shibamoto, Yasuteru; Hibiki, Takashi*; Tsukamoto, Naofumi*; Kaneko, Junichi*
JAEA-Review 2024-039, 45 Pages, 2024/09
JAEA-Review-2024-039.pdf:2.23MB
Several heat transfer correlations have been reported related to single-phase opposing flow; however, these correlations are based on experiments conducted in various channel geometries, working fluids, and thermal flow parameter ranges. Therefore, establishing a guideline for deciding which correlation should be selected based on its range of applicability and extrapolation performance is important. This study reviewed the existing heat transfer correlations for turbulent opposing-flow mixed convection. Furthermore, the authors evaluated the predictive performance of each correlation by comparing them with the experimental data obtained under various experimental conditions. The Jackson and Fewster, Churchill, and Swanson and Catton correlations can accurately predict all the experimental data. The authors confirmed that heat transfer correlations using the hydraulic-equivalent diameter as a characteristic length can be used for predictions regardless of channel-geometry differences. Furthermore, correlations described based on nondimensional dominant parameters can be used for predictions regardless of the differences in working fluids.
Pu
)O (x = 0, 0.18, 0.45, and 1) and analysis of heat capacityHirooka, Shun; Morimoto, Kyoichi; Matsumoto, Taku; Ogasawara, Masahiro*; Kato, Masato; Murakami, Tatsutoshi
Journal of Nuclear Materials, 598, p.155188_1 - 155188_9, 2024/09
Times Cited Count：0 Percentile：0.00(Materials Science, Multidisciplinary)no abstracts in English
Nagatsuka, Kentaro; Noguchi, Hiroki; Nagasumi, Satoru; Nomoto, Yasunobu; Shimizu, Atsushi; Sato, Hiroyuki; Nishihara, Tetsuo; Sakaba, Nariaki
Nuclear Engineering and Design, 425, p.113338_1 - 113338_11, 2024/08
Times Cited Count：6 Percentile：92.92(Nuclear Science & Technology)HTGR has a potential to contribute to decarbonization of hard-to-abate industries by supplying a large amount of hydrogen and high temperature heat or steam without carbon dioxide emission. JAEA has been conducting R&Ds for HTGR technologies with High Temperature Engineering Test Reactor (HTTR). This paper shows that HTTR's tests including the loss of core cooing test as a joint the OECD/NEA international research project and a HTTR heat application test plan which demonstrate hydrogen production by coupling the HTTR with a hydrogen production test facility. Additionally, aiming for operation start from the latter half of 2030s, the basic design of the HTGR demonstration reactor has been shown. The Japan's HTGR technology capabilities established by the HTTR project will be fully utilized for the construction of HTGR demonstration reactor.
Yamano, Hidemasa; Futagami, Satoshi; Higurashi, Koichi*
Proceedings of Advanced Reactor Safety (ARS 2024), p.121 - 130, 2024/08
This paper describes the application of safety design criteria (SDC) and safety design guideline (SDG) developed in the Generation-IV international forum on decay heat removal system (DHRS) enhancing reliability to sodium-cooled fast reactors (SFRs) recently designed in Japan.
CoSiYamauchi, Hiroki; Sari, D. P.*; Yasui, Yukio*; Sakakura, Terutoshi*; Kimura, Hiroyuki*; Nakao, Akiko*; Ohara, Takashi; Honda, Takashi*; Kodama, Katsuaki; Igawa, Naoki; et al.
Physical Review Research (Internet), 6(1), p.013144_1 - 013144_9, 2024/02
Konno, Chikara
Journal of Nuclear Science and Technology, 61(1), p.121 - 126, 2024/01
Times Cited Count：2 Percentile：37.43(Nuclear Science & Technology)The JENDL-4.0/HE neutron and proton ACE files were produced in 2017 and those of 22 nuclei for neutron and 25 nuclei for proton were bundled in the PHITS code. Recently it was found that the following five data in the JENDL-4.0/HE neutron and proton ACE files had any problems; ACE files for 
N and 
O, heating numbers, damage energy production cross sections, secondary neutron multiplicities and fission cross sections. Thus new JENDL-4.0/HE neutron and proton ACE files were produced with the problems fixed. This paper describes the problems and how to produce the new neutron and proton ACE files in detail.
Suzuki, Seiya; Nemoto, Yoshihiro*; Shiiki, Natsumi*; Nakayama, Yoshiko*; Takeguchi, Masaki*
Annalen der Physik, 535(9), p.2300122_1 - 2300122_12, 2023/09
Times Cited Count：0 Percentile：0.00(Physics, Multidisciplinary)Thwe Thwe, A.; Kadowaki, Satoshi; Nagaishi, Ryuji
Journal of Nuclear Science and Technology, 60(6), p.731 - 742, 2023/06
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)In this study, we performed numerical calculations of unsteady reaction flow considering detailed chemical reactions, investigated the unstable behavior of hydrogen-air dilute premixed flame due to intrinsic instability, and clarified the effects of unburned gas temperature and pressure. I made it. The unstable behavior of the flame in a wide space was simulated, and the burning rate of the cellular flame was obtained. Then, the effects of heat loss and flame scale on flame unstable behavior were investigated. The burning velocity of a planar flame increases as the unburned-gas temperature increases and it decreases as the unburned-gas pressure and heat loss increase. The normalized burning velocity increases when the pressure increases and heat loss becomes large, and it decreases when the temperature increases. This is because the high unburned-gas pressure and heat loss promote the unstable behavior and instability of flame.
Abe, Satoshi; Shibamoto, Yasuteru
Nuclear Engineering and Technology, 55(5), p.1742 - 1756, 2023/05
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)