Shell evolution of = 40 isotones towards Ca; First spectroscopy of Ti

Corts, M. L.*; Rodriguez, W.*; Doornenbal, P.*; Obertelli, A.*; Holt, J. D.*; Lenzi, S. M.*; Menndez, J.*; Nowacki, F.*; Ogata, Kazuyuki*; Poves, A.*; et al.

Physics Letters B, 800, p.135071_1 - 135071_7, 2020/01

https://doi.org/10.1016/j.physletb.2019.135071

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.

Comparative methodology between actual RCCS and downscaled heat-removal test facility

Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*

Annals of Nuclear Energy, 133, p.830 - 836, 2019/11

https://doi.org/10.1016/j.anucene.2019.07.025

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.

Emergence of a complete heavy-quark spin symmetry multiplet; Seven molecular pentaquarks in light of the latest LHCb analysis

Liu, M.-Z.*; Pan, Y.-W.*; Peng, F.-Z.*; Snchez-Snchez, M.*; Geng, L.-S.*; Hosaka, Atsushi; Valderrama, M. P.*

Physical Review Letters, 122(24), p.242001_1 - 242001_5, 2019/06

https://doi.org/10.1103/PhysRevLett.122.242001

Experimental study on heat removal performance of a new Reactor Cavity Cooling System (RCCS)

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.

Particle-based simulation of heat transfer behavior in EAGLE ID1 in-pile test

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.

Dirac surface state-modulated spin dynamics in a ferrimagnetic insulator at room temperature

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

https://doi.org/10.1126/sciadv.aas8660

The Zr()Zr cross section from the surrogate ratio method and its effect on -process nucleosynthesis

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

https://doi.org/10.3847/1538-4357/aa8c74

Effect of seawater on heat transfer without boiling in internally heated annulus

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

https://doi.org/10.3327/taesj.J15.024

no abstracts in English

Critical heat flux prediction for subcooled flow boiling in annulus

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%.

Measurement of void fraction distribution in air-water two-phase flow in a 44 rod bundle

Liu, W.; Jiao, L.; Nagatake, Taku; Shibata, Mitsuhiko; Komatsu, Masao*; Takase, Kazuyuki*; Yoshida, Hiroyuki

Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 10 Pages, 2016/10

To contribute the clarification of the Fukushima Daiichi Accident, Japan Atomic Energy Agency (JAEA) has been performed experiments to obtain void fraction distribution data, including detailed bubble information such as bubble velocity and size, in steam-water two-phase flow in rod bundle geometry under high pressure and high temperature condition, focusing on low flow rate at the core natural circulation flow condition after the reactor scram. In this research, experimental apparatus for measuring void fraction distribution in the 44 rod bundle was constructed. To measure the void fraction distribution under high pressure and high temperature condition (up to 2.8 MPa, 232 C), two wire mesh sensors (WMSs) were installed. To confirm the applicability of the installed WMSs and the measuring system for two-phase flow in rod bundle, experiments in air-water two-phase flow under atmospheric pressure and room temperature were performed. As a result, it was confirmed that the installed WMSs can be applicable to the two-phase flow in rod bundle. Measured results, such as instantaneous and time-averaged void fraction distribution in the rod bundle, average void fraction across the cross section of the flow channel, bubble length and velocity, were also reported.