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Satou, Akira; Wada, Yuki; Shibamoto, Yasuteru
Nuclear Engineering and Design, 437, p.114020_1 - 114020_14, 2025/06
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Post-boiling transition (post-BT) heat transfer is essential for analyzing the duration of surface dryout and peak cladding temperature during abnormal transients and accidents in light water reactors. The rewetting phenomenon is very important for evaluating the dryout duration. However, due to the lack of an experimental database on rewetting velocities under high flow and heat flux conditions, sufficient data for model development and validation do not exist. Therefore, a database on rewetting velocities caused by stepwise boundary condition changes under a wide range and multiple combination of thermal-hydraulic conditions was obtained using a single-tube experimental apparatus. Based on this database and the characteristics of rewetting velocities obtained, an experimental correlation for rewetting velocity was proposed. This correlation predicts the rewetting velocity accurately by taking the change in the mass flux of the liquid or gas phase with stepwise transients as a parameter. This suggested that the change in the mass flux of the gas or liquid phase near the liquid film front has a strong influence on the rewetting under extremely high mass flux conditions compared to the reflooding process.
Onishi, Takashi; Koyama, Shinichi*; Yokoyama, Keisuke; Morishita, Kazuki; Watanabe, Masashi; Maeda, Shigetaka; Yano, Yasuhide; Oki, Shigeo
Nuclear Engineering and Design, 432, p.113755_1 - 113755_17, 2025/02
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Hamase, Erina; Miyake, Yasuhiro*; Imai, Yasutomo*; Doda, Norihiro; Ono, Ayako; Tanaka, Masaaki
Nuclear Engineering and Design, 432, p.113738_1 - 113738_12, 2025/02
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)To enhance the safety of sodium-cooled fast reactors, the natural circulation (NC) decay heat removal systems with a dipped-type direct heat exchanger (D-DHX) have been investigated. During the D-DHX operation, since the core-plenum interaction occurs, the reactor vessel model using a computational fluid dynamics code (RV-CFD) is required to be established. Previously, the CFD model based on the subchannel analysis was developed. In this study, to achieve lower computational cost maintaining the prediction accuracy, the coarse-mesh subchannel CFD (CMSC) model was developed, and was incorporated into the core of RV-CFD. As a result of PLANDTL-1 test analysis, the RV-CFD with the CMSC model can reproduce the core-plenum interaction under NC conditions.
Matsumura, Taichi; Okumura, Keisuke; Sakamoto, Masahiro; Terashima, Kenichi; Riyana, E. S.; Kondo, Kazuhiro*
Nuclear Engineering and Design, 432, p.113791_1 - 113791_9, 2025/02
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Matsushita, Kentaro; Ezure, Toshiki; Tanaka, Masaaki; Imai, Yasutomo*; Fujisaki, Tatsuya*; Sakai, Takaaki*
Nuclear Engineering and Design, 432, p.113785_1 - 113785_16, 2025/02
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Establishing an evaluation method for the gas entrainment (GE) of argon cover gas due to surface vortices is required in terms of safety design of sodium-cooled fast reactors. To modify the evaluation model in an in-house evaluation tool for GE, StreamViewer, a modified evaluation model on the pressure distribution along the vortex center line (PVL model) was proposed to identify the vortex center lines by connecting continuous vortex center points from the suction port to the surface and evaluate gas core length based on the balance between the hydrostatic pressure and the pressure decrease distribution along the vortex center line. PVL model was applied the three-dimensional numerical analysis results for the experiments where a plate induced unsteady traveling vortices in the open channel flow. Consequently, the GE evaluation using StreamViewer with PVL model could reproduce the relation between the inlet flow velocity and the gas core length in the unsteady vortex flow experiments.
Yoshida, Masato; Iguchi, Satoshi; Hirano, Hiroshi*; Kitamura, Akihiro
Nuclear Engineering and Design, 431, p.113691_1 - 113691_16, 2025/01
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)The Plutonium Fuel Fabrication Facility is currently in the decommissioning phase, with glovebox dismantling operations ongoing since 2010. During conventional glovebox dismantling operations, the glovebox to be dismantled is enclosed within plastic tents to contain contamination. The glovebox is then dismantled by workers wearing air-fed suits with thermal or mechanical cutting tools, which typically generate dross or sparks in the form of radioactive aerosols during cutting. Despite the longevity and meticulous organization of this manual method, the workload remains considerable, while the allowable working time is limited. In addition, the potential for inhalation exposure to plutonium is elevated in the event of an accident given the contamination of the work area. To overcome disadvantages associated with conventional glovebox dismantling methods, new methods are currently being developed. The primary objective is to reduce the reliance on operation based on air-fed suits and enhance worker safety by introducing remote equipment and a new floor-reinforcing panel. Another objective is to suppress waste generation by reusing all equipment on multiple occasions which is achieved by developing a containment system that have a large open port with a pallet for the storage and reuse of equipment for successive operations. Furthermore, a glove operation compartment is designed and tested for the manual handling of dismantled materials as an additional strategy to reduce work based on air-fed suits and mitigate secondary waste generation.
at temperature range 170 - 290
CLuu, V. N.; Nakajima, Kunihisa
Nuclear Engineering and Design, 426, p.113402_1 - 113402_7, 2024/09
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)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：4 Percentile：93.24(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.
Johnson, M.*; Emura, Yuki; Clavier, R.*; Matsuba, Kenichi; Kamiyama, Kenji; Brayer, C.*; Journeau, C.*
Nuclear Engineering and Design, 423, p.113165_1 - 113165_14, 2024/07
Times Cited Count：2 Percentile：57.00(Nuclear Science & Technology)Experimental investigation of two interactions between molten jets and sodium, pertaining to severe accidents in a sodium-cooled fast reactor, have been undertaken at the JAEA's MELT facility. X-ray imaging and debris analysis reveal rapid formation of a crust at the melt coolant-interface, instigating thermal fragmentation events. Heat transfer calculations at the jet-coolant interface, supported by particle tracking velocimetry characterisation of the jet velocity, imply the formation of a solid crust within milliseconds of contact with the coolant. A mechanism for enhanced thermal fragmentation is proposed, inspired by observations from the X-ray imaging of coolant entrainment into the jet.
compoundsFrazer, D.*; Saleh, T. A.*; Matsumoto, Taku; Hirooka, Shun; Kato, Masato; McClellan, K.*; White, J. T.*
Nuclear Engineering and Design, 423, p.113136_1 - 113136_7, 2024/07
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Nanoindentation based techniques can be employed on minute volumes of material to measure mechanical properties, including Young's modulus, hardness, and creep stress exponents. In this study, (U,Ce)O solid solutions samples are used to develop elevated temperature nanoindentation and nanoindentation creep testing methods for use on mixed oxide fuels. Nanoindentation testing was performed on 3 separate (Ux-1,Cex)O compounds ranging from x equals 0.1 to 0.3 at up to 800 C: their Young's modulus, hardness, and creep stress exponents were evaluated. The Young's modulus decreases in the expected linear manner while the hardness decreases in the expected exponential manner. The nanoindentation creep experiments at 800 C give stress exponent values, n=4.7-6.9, that suggests dislocation motion as the deformation mechanism.
Sato, Ikken; Yoshikawa, Shinji; Yamashita, Takuya; Shimomura, Kenta; Cibula, M.*; Mizokami, Shinya*
Nuclear Engineering and Design, 422, p.113088_1 - 113088_24, 2024/06
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Kamide, Hideki; Asayama, Tai; Wakai, Takashi; Ezure, Toshiki; Uchibori, Akihiro; Kubo, Shigenobu; Takeuchi, Masayuki
Nuclear Engineering and Design, 421, p.113062_1 - 113062_10, 2024/05
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)A sodium cooled fast reactor (SFR) is one of the most relevant and decarbonized energy supply system with higher sustainability on natural resources, footprint, and waste management. It was planned in a strategic roadmap of fast reactor decided by Inter-Ministerial Council for Nuclear Power Japan in 2022 to start a conceptual design of a demonstration reactor from 2024 with a background of accumulated knowledge and experiences of SFR development. For example, a design and lifecycle simulation/evaluation system named ARKADIA has been developed to accelerate such design works. It will enable to take into account plant lifecycle, e.g., operation and maintenance, to the plant design and optimize it based on simulations and knowledgebase. This paper shows research progresses of ARKADIA, safety design and evaluations, codes and standards, fuel cycle, and SFR development projects in Japan.
Baccou, J.*; Glantz, T.*; Ghione, A.*; Sargentini, L.*; Fillion, P.*; Damblin, G.*; Sueur, R.*; Iooss, B.*; Fang, J.*; Liu, J.*; et al.
Nuclear Engineering and Design, 421, p.113035_1 - 113035_16, 2024/05
Times Cited Count：6 Percentile：97.32(Nuclear Science & Technology)Ho, H. Q.; Ishii, Toshiaki; Nagasumi, Satoru; Ono, Masato; Shimazaki, Yosuke; Ishitsuka, Etsuo; Sawahata, Hiroaki; Goto, Minoru; Simanullang, I. L.*; Fujimoto, Nozomu*; et al.
Nuclear Engineering and Design, 417, p.112795_1 - 112795_6, 2024/02
Times Cited Count：1 Percentile：0.00(Nuclear Science & Technology)Soma, Shu; Ishigaki, Masahiro*; Abe, Satoshi; Shibamoto, Yasuteru
Nuclear Engineering and Design, 416, p.112754_1 - 112754_18, 2024/01
Times Cited Count：1 Percentile：25.62(Nuclear Science & Technology)Pshenichnikov, A.; Nagae, Yuji
Nuclear Engineering and Design, 415, p.112729_1 - 112729_16, 2023/12
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Sato, Ikken; Yoshikawa, Shinji; Yamashita, Takuya; Shimomura, Kenta; Cibula, M.*; Mizokami, Shinya*
Nuclear Engineering and Design, 414, p.112574_1 - 112574_20, 2023/12
Uchibori, Akihiro; Doda, Norihiro; Aoyagi, Mitsuhiro; Sonehara, Masateru; Sogabe, Joji; Okano, Yasushi; Takata, Takashi*; Tanaka, Masaaki; Enuma, Yasuhiro; Wakai, Takashi; et al.
Nuclear Engineering and Design, 413, p.112492_1 - 112492_10, 2023/11
Times Cited Count：2 Percentile：46.61(Nuclear Science & Technology)The ARKAIDA has been developed to realize automatic optimization of plant design from safety evaluation for the advanced reactors represented by a sodium-cooled fast reactor. ARKADIA-Design offers functions to support design optimization both in normal operating conditions and design basis events. The multi-level simulation approach by the coupled analysis such as neutronics, core deformation, core thermal hydraulics was developed as one of the main technologies. On the other hand, ARKAIDA-Safety aims for safety evaluation considering severe accidents. As a key technology, the numerical methods for in- and ex-vessel coupled phenomena during severe accidents in sodium-cooled fast reactors were tested through a hypothetical severe accident event. Improvement of the ex-vessel model and development of the AI technology to find best design solution have been started.
Kitamura, Akihiro; Hirano, Hiroshi*; Yoshida, Masato
Nuclear Engineering and Design, 411, p.112435_1 - 112435_14, 2023/09
Times Cited Count：1 Percentile：25.62(Nuclear Science & Technology)This study presents the features and brief history of the glovebox dismantling facility and the primary dismantlement results. Subsequently, we evaluate the novelties of the facility from operational experiences in manual and remote glovebox dismantlement methods and discuss their characteristics. Furthermore, we evaluate the worker exposure dose based on the obtained data. Finally, we show how these experiences are effectively fed back to the technological dismantlement development for our decommissioning project.
Narukawa, Takafumi; Hamaguchi, Shusuke*; Takata, Takashi*; Udagawa, Yutaka
Nuclear Engineering and Design, 411, p.112443_1 - 112443_12, 2023/09
Times Cited Count：1 Percentile：25.62(Nuclear Science & Technology)