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Sonehara, Masateru; Okano, Yasushi; Uchibori, Akihiro; Oki, Hiroshi*
Journal of Nuclear Science and Technology, 62(5), p.403 - 414, 2025/05
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)For sodium-cooled fast reactors, understanding sodium combustion behaviour is crucial for managing sodium leakage accidents. In this study, we perform benchmark analyses of the Sandia National Laboratories (SNL) T3 experiment using the multi-dimensional thermal hydraulic code AQUA-SF. Conducted in an enclosed space with a large vessel volume of 100 m
and a sodium mass flow rate of 1 kg/s, the experiment highlighted the multi-dimensional effects of local temperature increase shortly after sodium injection. This study aims to extend the capabilities of AQUA-SF by focusing on the simulation of these multi-dimensional temperature variations, in particular the formation of high temperature regions at the bottom of the vessel. The proposed models include the temporary stopping of sodium droplet ignition and spray combustion of sodium splash on the floor. Furthermore, it has been shown that additional heat source near the floor is essential to enhance the reproduction of the high temperature region at the bottom. Therefore, case studies including sensitivity analyses of spray cone angle and prolonged combustion of droplets on the floor are conducted. This comprehensive approach provides valuable insights into the dynamics of sodium combustion and safety measures in sodium-cooled fast reactors.
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：1 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.
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)Calabrese, R.*; Hirooka, Shun
Progress in Nuclear Energy, 178, p.105516_1 - 105516_11, 2025/01
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Thermal creep is one of the key properties of mixed oxide (MOX) fuel for innovative fast reactors. Thermal creep of fuel affects markedly the interaction between the fuel and the cladding. A review of correlations available in the literature is presented. The effect of porosity, plutonium concentration, and stoichiometry are discussed also in the light of recent numerical results. Our analysis pointed out some inconsistencies concerning the modelling of the effect of porosity on diffusional creep and a re-evaluation of the effect of plutonium concentration. The discussion suggested that Evans's findings on the effect of stoichiometry should be better assessed as well as the level of increase in creep moving towards stoichiometry. Typical operating conditions of fast breeder reactors confirmed the need for an extension of porosity and temperature correlations' domains. Besides this, a new correlation based on a separate-effect approach has been proposed for fuel performance codes.
Doi, Daisuke
International Journal of Hydrogen Energy, 91, p.1245 - 1252, 2024/11
Times Cited Count：1 Percentile：25.55(Chemistry, Physical)Emura, Yuki; Matsuba, Kenichi; Kikuchi, Shin; Yamano, Hidemasa
Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (Internet), 8 Pages, 2024/11
Miyazawa, Takeshi; Uwaba, Tomoyuki; Yano, Yasuhide; Tanno, Takashi; Otsuka, Satoshi; Onizawa, Takashi; Ando, Masanori; Kaito, Takeji
JAEA-Technology 2024-009, 140 Pages, 2024/10
JAEA-Technology-2024-009.pdf:8.03MB
For the purpose of enhancing the reliability of fast reactor fuel designing using modified type 316 steel, the out-of-pile and in-pile mechanical data of modified type 316 steel cladding tubes and wrapper tubes were statistically analyzed with the knowledge on material science and engineering; the high-temperature strength equations of modified type 316 steel, which can be applied to high-dose neutron irradiation environment, were derived. The out-of-pile high-temperature tensile and creep data of modified type 316 steel cladding tubes and wrapper tubes were derived up to 900
C, which is higher than the upper limit temperature of anticipated transient event of fast reactor. Using the extended database, the best-fit equation and the lower limit equation were derived for out-of-pile 0.2% proof strength, ultimate tensile strength and creep rupture strength while the best-fit equation and the upper and lower limit equations for creep strain. Furthermore, the degradation factors for tensile and creep strength, which will be produced by in-reactor environment (i.e., neutron irradiation in liquid sodium), were evaluated using the existing neutron irradiation data of modified type 316 steel, which were derived using the experimental fast reactor Joyo, the French proto-type fast reactor Phenix, the American experimental fast reactor FFTF. The derived equations were validated by the comparison with the experimental data.
Yokoyama, Kenji; Hazama, Taira; Taninaka, Hiroshi; Oki, Shigeo
JAEA-Data/Code 2024-007, 41 Pages, 2024/10
JAEA-Data-Code-2024-007.pdf:1.1MB
The third version of the versatile reactor analysis code system, MARBLE3, has been developed. In the development of the former version of MARBLE, object-oriented scripting language Python (Python2) had been used and then the latest version of Python (Python3) was released. However, due to its backward incompatibility, MARBLE no longer worked with Python3. For this reason, MARBLE3 has been fully modified and maintained to work with Python3. In MARBLE3, newly developed analysis codes and newly proposed calculation methods were incorporated, and the user interface was extended and solvers were reimplemented for maintainability, extensibility, and flexibility. In MARBLE3, the three-dimensional hexagonal/triangular transport code MINISTRI Ver.7 (MINISTRI7) and the three-dimensional hexagonal/triangular diffusion code D-MINISTRI are available as the new analysis codes. These codes can be used in the neutronics analysis system SCHEME and the fast reactor burnup analysis system OPRHEUS, which are the subsystems of MARBLE. In addition, the user interface of CBG, a core analysis system embedded in MARBLE, was extended so that the diffusion and transport calculation solvers for the 2-dimensional RZ system of CBG can be used on SCHEME. On the other hand, MARBLE3 has extended the functionality of the burnup calculation solver so that it can use the numerical methods proposed in the papers on the improvement of the Chebyshev rational function approximation method and the minimax polynomial approximation method. From the viewpoint of maintainability, the point reactor kinetics solver POINTKINETICS, which was introduced in MARBLE2, has been newly reworked as the KINETICS solver in MARBLE3.
Onoda, Yuichi; Ishida, Shinya; Fukano, Yoshitaka; Kamiyama, Kenji; Yamano, Hidemasa; Kubo, Shigenobu; Shibata, Akihiro*; Bertrand, F.*; Seiler, N.*
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
Sogabe, Joji; Ishida, Shinya; Tagami, Hirotaka; Okano, Yasushi; Kamiyama, Kenji; Onoda, Yuichi; Matsuba, Kenichi; Yamano, Hidemasa; Kubo, Shigenobu; Kubota, Ryuzaburo*; et al.
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
In the frame of France-Japan collaboration, the calculational methodologies were defined and assessed, and the phenomenology and the severe accident consequences were investigated in a pool-type sodium-cooled fast reactor.
Kurisaka, Kenichi; Nishino, Hiroyuki; Yamano, Hidemasa
Proceedings of Probabilistic Safety Assessment and Management & Asian Symposium on Risk Assessment and Management (PSAM17 & ASRAM2024) (Internet), 10 Pages, 2024/10
The objective of this study is to implement an effectiveness evaluation of the measures for improving resilience of nuclear structures against excessive earthquake. In this study, those measures for improving resilience have an effect to enlarge their seismic safety margin. To evaluate effectiveness of those measures, seismic core damage frequency (CDF) is selected as an index. Reduction of CDF as an effectiveness index is quantified by applying seismic PRA technology. Target system is a loop-type next-generation sodium-cooled fast reactor, which adopts the building isolated from horizontal seismic ground motion. Even if the reactor vessel (RV) is buckled due to seismic shaking, it is expected that the RV maintains stable state without unstable failure such as rupture, collapse. Realistic consideration of the post-buckling behavior is regarded as a measure for improving resilience in this study. We set two cases for improving the resilience in the accident sequences analysis. The first case assumes low-cycle fatigue failure after buckling as the realistic failure mode of the RV, and we applied the fragility evaluated in our study. After the RV fatigue failure, the behavior of failure propagation is very uncertain. As the second case, the median seismic capacity to loss of reactor level is assumed to be slightly larger than that of fatigue failure of the RV. Analyses for both cases were performed, and the results were compared to the base case indicating significant reduction of CDF. Within the assumption, the measures for improving the resilience were significantly effective for decreasing CDF in excessive earthquake up to several times of a design basis ground motion. The seismic PRA technology could serve to the effectiveness evaluation of the measures for improving resilience of nuclear structures against excessive earthquake.
Yamano, Hidemasa; Futagami, Satoshi; Shibata, Akihiro*
Proceedings of Advanced Reactor Safety (ARS 2024), p.151 - 160, 2024/08
This study examined the application of safety design criteria (SDC) and safety design guideline (SDG) developed in the Generation-IV international forum on the active reactor shutdown system (RSS) to sodium-cooled fast reactors (SFRs) recently designed in Japan.
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.
Pu
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powder using master sintering curve theoryNakamichi, Shinya; Sunaoshi, Takeo*; Hirooka, Shun; Vauchy, R.; Murakami, Tatsutoshi
Journal of Nuclear Materials, 595, p.155072_1 - 155072_11, 2024/07
Times Cited Count：1 Percentile：51.66(Materials Science, Multidisciplinary)Li, C.-Y.; Watanabe, Akira*; Uchibori, Akihiro; Okano, Yasushi
Journal of Nuclear Science and Technology, 61(7), p.935 - 957, 2024/07
Times Cited Count：2 Percentile：43.92(Nuclear Science & Technology)Emura, Yuki; Takai, Toshihide; Kikuchi, Shin; Kamiyama, Kenji; Yamano, Hidemasa; Yokoyama, Hiroki*; Sakamoto, Kan*
Journal of Nuclear Science and Technology, 61(7), p.911 - 920, 2024/07
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Ishida, Shinya; Uchibori, Akihiro; Okano, Yasushi
Dai-28-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2024/06
no abstracts in English
Miyazawa, Takeshi; Tanno, Takashi; Imagawa, Yuya; Hashidate, Ryuta; Yano, Yasuhide; Kaito, Takeji; Otsuka, Satoshi; Mitsuhara, Masatoshi*; Toyama, Takeshi*; Onuma, Masato*; et al.
Journal of Nuclear Materials, 593, p.155008_1 - 155008_16, 2024/05
Times Cited Count：1 Percentile：51.66(Materials Science, Multidisciplinary)Ishida, Shinya; Fukano, Yoshitaka; Tobita, Yoshiharu; Okano, Yasushi
Journal of Nuclear Science and Technology, 61(5), p.582 - 594, 2024/05
Times Cited Count：1 Percentile：23.64(Nuclear Science & Technology)Yoshikawa, Ryuji; Imai, Yasutomo*; Kikuchi, Norihiro; Tanaka, Masaaki; Ohshima, Hiroyuki
Nuclear Technology, 210(5), p.814 - 835, 2024/05
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)In the study of safety enhancement on advanced sodium-cooled fast reactor, it is essential to clarify the thermal-hydraulics under various operation conditions in a fuel assembly (FA) with the wire-wrapped fuel pins to assess the structural integrity of the fuel pin. A finite element thermal-hydraulics analysis code named SPIRAL has been developed to analyze the detailed thermal-hydraulics phenomena in a FA. In this study, the numerical simulations of the 37-pin bundle sodium experiments at different Re number conditions, including a transitional condition between laminar and turbulent flows and turbulent flow conditions, were performed to validate the hybrid turbulence model equipped in SPIRAL. The temperature distributions predicted by SPIRAL was consistent with those measured in the experiments. Through the validation study, the applicability of the hybrid turbulence model in SPIRAL to thermal-hydraulic evaluation of sodium-cooled FA in the wide range of Re number was confirmed.
