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Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*
Annals of Nuclear Energy, 151, p.107867_1 - 107867_11, 2021/02
A new RCCS with passive safety features consists of two continuous closed regions. One is a region surrounding RPV. The other is a cooling region with heat transferred to the ambient air. The new RCCS needs no electrical or mechanical driving devices. We compared the RCCS using atmospheric radiation with that using atmospheric natural circulation in terms of passive safety features and control methods for heat removal. The magnitude relationship for passive safety features is heat conduction radiation
natural convection. Therefore, the magnitude for passive safety features of the former RCCS can be higher than that of the latter RCCS. In controlling the heat removal, the former RCCS changes the heat transfer area only. On the other hand, the latter RCCS needs to change the chimney effect. It is necessary to change the air resistance in the duct. Therefore, the former RCCS can control the heat removal more easily than the latter RCCS.
Zhang, T.*; Funakoshi, Kanji*; Liu, X.*; Liu, W.*; Morita, Koji*; Kamiyama, Kenji
Annals of Nuclear Energy, 150, p.107856_1 - 107856_10, 2021/01
Corts, M. L.*; Rodriguez, W.*; Doornenbal, P.*; Obertelli, A.*; Holt, J. D.*; Lenzi, S. M.*; Men
ndez, J.*; Nowacki, F.*; Ogata, Kazuyuki*; Poves, A.*; et al.
Physics Letters B, 800, p.135071_1 - 135071_7, 2020/01
Times Cited Count:3 Percentile:13.44(Astronomy & Astrophysics)Excited states in the = 40 isotone
Ti were populated via the
V(
,
)
Ti reaction at
200 MeV/nucleon at the Radioactive Isotope Beam Factory and studied using
-ray spectroscopy. The energies of the
and
transitions, observed here for the first time, indicate a deformed Ti ground state. These energies are increased compared to the neighboring
Cr and
Fe isotones, suggesting a small decrease of quadrupole collectivity. The present measurement is well reproduced by large-scale shell-model calculations based on effective interactions, while ab initio and beyond mean-field calculations do not yet reproduce our findings.
Sun, M. D.*; Liu, Z.*; Huang, T. H.*; Zhang, W. Q.*; Andreyev, A. N.; Ding, B.*; Wang, J. G.*; Liu, X. Y.*; Lu, H. Y.*; Hou, D. S.*; et al.
Physics Letters B, 800, p.135096_1 - 135096_5, 2020/01
Times Cited Count:1 Percentile:30.54(Astronomy & Astrophysics)Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*
Annals of Nuclear Energy, 133, p.830 - 836, 2019/11
Times Cited Count:0 Percentile:100(Nuclear Science & Technology)A RCCS having passive safety features through radiation and natural convection was proposed. The RCCS design consists of two continuous closed regions: an ex-reactor pressure vessel region and a cooling region with a heat-transfer surface to ambient air. The RCCS uses a novel shape to remove efficiently the heat released from the RPV through as much radiation as possible. Employing air as the working fluid and ambient air as the ultimate heat sink, the RCCS design can strongly reduce the possibility of losing the working fluid and the heat sink for decay-heat-removal. Moreover, the authors started experiment research with using a scaled-down heat-removal test facility. Therefore, this study propose a comparative methodology between an actual RCCS and a scaled-down heat-removal test facility.
Chen, S.*; Lee, J.*; Doornenbal, P.*; Obertelli, A.*; Barbieri, C.*; Chazono, Yoshiki*; Navrtil, P.*; Ogata, Kazuyuki*; Otsuka, Takaharu*; Raimondi, F.*; et al.
Physical Review Letters, 123(14), p.142501_1 - 142501_7, 2019/10
Times Cited Count:8 Percentile:16.18(Physics, Multidisciplinary)no abstracts in English
Liu, M.-Z.*; Pan, Y.-W.*; Peng, F.-Z.*; Snchez-S
nchez, M.*; Geng, L.-S.*; Hosaka, Atsushi; Valderrama, M. P.*
Physical Review Letters, 122(24), p.242001_1 - 242001_5, 2019/06
Times Cited Count:68 Percentile:0.22(Physics, Multidisciplinary)Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*
Annals of Nuclear Energy, 122, p.201 - 206, 2018/12
Times Cited Count:1 Percentile:75.54(Nuclear Science & Technology)A RCCS having passive safety features through radiation and natural convection was proposed. The RCCS design consists of two continuous closed regions: an ex-reactor pressure vessel region and a cooling region with a heat-transfer surface to ambient air. The RCCS uses a novel shape to remove efficiently the heat released from the RPV through as much radiation as possible. Employing air as the working fluid and ambient air as the ultimate heat sink, the RCCS design can strongly reduce the possibility of losing the working fluid and the heat sink for decay-heat-removal. This study addresses an improvement of heat-removal capability using heat conduction on the RCCS. As a result, a heat flux removed by the RCCS could be doubled; therefore, it is possible to halve the height of the RCCS or increase the thermal reactor power.
Hosomi, Seisuke*; Akashi, Tomoyasu*; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*; Takamatsu, Kuniyoshi
Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 7 Pages, 2018/11
A new RCCS with passive safety features consists of two continuous closed regions. One is a region surrounding RPV. The other is a cooling region with heat transferred to the ambient air. The new RCCS needs no electrical or mechanical driving devices. We started experiment research with using a scaled-down test section. Three experimental cases under different emissivity conditions were performed. We used Monte Carlo method to evaluate the contribution of radiation to the total heat released from the heater. As a result, after the heater wall was painted black, the contribution of radiation to the total heat could be increased to about 60%. A high emissivity of RPV surface is very effective to remove more heat from the reactor. A high emissivity of the cooling part wall is also effective because it not only increases the radiation emitted to the ambient air, but also may increase the temperature difference among the walls and enhance the convection heat transfer in the RCCS.
Kato, Masatsugu*; Funakoshi, Kanji*; Liu, X.*; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*; Kamiyama, Kenji
Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 7 Pages, 2018/11
Morita, Koji*; Ogawa, Ryusei*; Tokioka, Hiromi*; Liu, X.*; Liu, W.*; Kamiyama, Kenji
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 11 Pages, 2018/10
The EAGLE in-pile ID1 test has been performed by Japan Atomic Energy Agency to demonstrate early fuel discharge from a fuel subassembly with an inner duct structure, which is named FAIDUS. It was deduced that early duct wall failure observed in the test was initiated by high heat flux from the molten pool of fuel and steel mixture. The posttest analyses suggest that molten pool-to-duct wall heat transfer might be enhanced effectively by the molten steel with large thermal conductivity in the pool without the presence of fuel crust on the duct wall. In this study, mechanisms of heat transfer from the molten pool to the duct wall was analyzed using a fully Lagrangian approach based on the finite volume particle method for multi-component, multi-phase flows. A series of pin disruption, molten pool formation and duct wall failure behaviors was simulated to investigate mixing and separation behavior of molten steel and fuel in the pool, and their effect on molten pool-to-duct wall heat transfer. The present 2D particle-based simulations demonstrated that large thermal load beyond 10 MW/m on the duct wall was caused by effective heat transfer due to direct contact of liquid fuel with nuclear heat to the duct wall.
Phan, L. H. S.*; Ohara, Yohei*; Kawata, Ryo*; Liu, X.*; Liu, W.*; Morita, Koji*; Guo, L.*; Kamiyama, Kenji; Tagami, Hirotaka
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 12 Pages, 2018/10
Self-leveling behavior of core fuel debris beds is one of the key phenomena for the safety assessment of core disruptive accidents (CDAs) in sodium-cooled fast reactors (SFRs). The SIMMER code has been developed for CDA analysis of SFRs, and the code has been successfully applied to numerical simulations for key thermal-hydraulic phenomena involved in CDAs as well as reactor safety assessment. However, in SIMMER's fluid-dynamics model, it is always difficult to represent the strong interactions between solid particles as well as the discrete particle characteristics. To solve this problem, a new method has been developed by combining the multi-fluid model of the SIMMER code with the discrete element method (DEM) for the solid phase to reasonably simulate the particle behaviors as well as the fluid-particle interactions in multi-phase flows. In this study, in order to validate the multi-fluid model of the SIMMER code coupled with DEM, numerical simulations were performed on a series of self-leveling experiments using a gas injection method in cylindrical particle beds. The effects of friction coefficient on the simulation results were investigated by sensitivity analysis. Though more extensive validations are needed, the reasonable agreement between simulation results and corresponding experimental data preliminarily demonstrates the potential ability of the present method in simulating the self-leveling behaviors of debris bed. It is expected that the SIMMER code coupled with DEM is a prospective computational tool for analysis of safety issues related to solid particle debris bed in SFRs.
Tang, C.*; Song, Q.*; Chang, C.-Z.*; Xu, Y.*; Onuma, Yuichi; Matsuo, Mamoru*; Liu, Y.*; Yuan, W.*; Yao, Y.*; Moodera, J. S.*; et al.
Science Advances (Internet), 4(6), p.eaas8660_1 - eaas8660_6, 2018/06
Times Cited Count:8 Percentile:26.78(Multidisciplinary Sciences)Chadwick, M. B.*; Capote, R.*; Trkov, A.*; Herman, M. W.*; Brown, D. A.*; Hale, G. M.*; Kahler, A. C.*; Talou, P.*; Plompen, A. J.*; Schillebeeckx, P.*; et al.
Nuclear Data Sheets, 148, p.189 - 213, 2018/02
Times Cited Count:22 Percentile:3.55(Physics, Nuclear)The CIELO collaboration has studied neutron cross sections on nuclides that significantly impact criticality in nuclear facilities - U,
U,
Pu,
Fe,
O and
H - with the aim of improving the accuracy of the data and resolving previous discrepancies in our understanding. This multi-laboratory pilot project, coordinated via the OECD/NEA Working Party on Evaluation Cooperation (WPEC) Subgroup 40 with support also from the IAEA, has motivated experimental and theoretical work and led to suites of new evaluated libraries that accurately reflect measured data and also perform well in integral simulations of criticality. This report summarizes our results and outlines plans for the next phase of this collaboration.
Yan, S. Q.*; Li, Z. H.*; Wang, Y. B.*; Nishio, Katsuhisa; Lugaro, M.*; Karakas, A. I.*; Makii, Hiroyuki; Mohr, P.*; Su, J.*; Li, Y. J.*; et al.
Astrophysical Journal, 848(2), p.98_1 - 98_8, 2017/10
Times Cited Count:0 Percentile:100(Astronomy & Astrophysics)Tam, D. M.*; Song, Y.*; Man, H.*; Cheung, S. C.*; Yin, Z.*; Lu, X.*; Wang, W.*; Frandsen, B. A.*; Liu, L.*; Gong, Z.*; et al.
Physical Review B, 95(6), p.060505_1 - 060505_6, 2017/02
Times Cited Count:15 Percentile:23.77(Materials Science, Multidisciplinary)Uesawa, Shinichiro; Liu, W.; Jiao, L.; Nagatake, Taku; Takase, Kazuyuki; Shibata, Mitsuhiko; Yoshida, Hiroyuki
Nippon Genshiryoku Gakkai Wabun Rombunshi, 15(4), p.183 - 191, 2016/12
no abstracts in English
Liu, W.; Nagatake, Taku; Shibata, Mitsuhiko; Koizumi, Yasuo; Yoshida, Hiroyuki; Nemoto, Yoshiyuki; Kaji, Yoshiyuki
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 4 Pages, 2016/11
The Fukushima Daiichi NPP accident asks that the accident management of the LOCA in the SFPs must be considered to avoid occurrences of severe accident in the SFPs. To prevent the failure of the spent fuel assemblies at the LOCA, transportable spray systems are expected to be put into use to discharge water into fuel assemblies to moderate the temperature increase. To apply the spray system as a countermeasure for the LOCA of the SFP, the capability of the spray cooling system must be evaluated to keep the spent fuel rods safety. JAEA has started the research project to investigate the spray cooling capability for the SFP. In this research project, we aim to construct a numerical simulation method for evaluating the capability of the spray cooling. To develop the method, the basic key phenomena that affect the cooling performance must be clarified and the validation data required for the code development. To clarify the basic key phenomena that affect the cooling performance, that is, the CCFL and the drop size effect on the CCFL, and to obtain the code validation data, we are planning to carry out 2 experiments with two test sections, the spray visualization experiment and the spray cooling experiment. The spray visualization test section aims to get CCFL data in air-water two-phase flow and to understand the two-phase flow behavior over the upper tie plate. The spray cooling test section aims to get the CCFL data in steam-water two-phase flow and to obtain the validation data. This paper focus on the outline of the research plan for the whole research project.
Liu, W.; Podowski, M. Z.*
Nippon Kikai Gakkai Netsu Kogaku Konfarensu 2016 Koen Rombunshu (USB Flash Drive), 2 Pages, 2016/10
Prediction of Critical Heat Flux (CHF) is important for nuclear reactor safety. However, the CHF prediction for subcooled flow boiling in complicated geometry such as fuel assembly still remains unsolved. As the first step for the CHF prediction in rod bundles, in this paper, we tried to predict the CHF in annulus, which is the most basic flow geometry simplified from a fuel bundle. We performed the CHF prediction by using liquid sublayer dryout model, combining with ANSYS CFX code to get the single phase velocity distribution inside the annulus. The results show that the CHF in annulus can be predicted in an accuracy of about 20%.
Jiao, L.; Liu, W.; Nagatake, Taku; Uesawa, Shinichiro; Shibata, Mitsuhiko; Yoshida, Hiroyuki; Takase, Kazuyuki*
Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 11 Pages, 2016/10