Paramagnetic electron-nuclear spin entanglement in HoCoZn

Kitazawa, Takafumi; Shimura, Yasuyuki*; Onimaru, Takahiro*; Tsuchida, Shun*; Kubo, Katsunori; Haga, Yoshinori; Sakai, Hironori; Tokiwa, Yoshifumi; Kambe, Shinsaku; Tokunaga, Yo

Physical Review Research (Internet), 8(2), p.023009_1 - 023009_14, 2026/04

https://doi.org/10.1103/bzvk-grrv

Aerial monitoring around TEPCO's Fukushima Daiichi Nuclear Power Station and development of radiation monitoring technology for unmanned vehicle in fiscal year 2024 (Contract research)

Futemma, Akira; Ochi, Kotaro; Sasaki, Miyuki; Nakama, Shigeo; Kawasaki, Yoshiharu*; Iwai, Takeyuki*; Hiraga, Shogo*; Haginoya, Masashi*; Matsunaga, Yuki*; Yamada, Tsutomu*; et al.

JAEA-Technology 2025-016, 253 Pages, 2026/03

JAEA-Technology-2025-016.pdf:20.16MB

Aerial Radiation Monitoring (ARM) has been used to quickly and widely measure radiation distribution caused by the TEPCO's Fukushima Daiichi Nuclear Power Station (FDNPS) accident resulted from the tsunami accompanying the Pacific coast of Tohoku Earthquake on March 11, 2011. Since the accident, As a commissioned project of the Nuclear Regulation Authority, the Japan Atomic Energy Agency (JAEA) has continuously conducted ARM around FDNPS. This report summarizes the results of the 2024 monitoring activities, evaluates temporal changes in ambient dose rates, and identifies factors contributing to these changes. A terrain-corrected analysis was applied to improve dose rates conversion accuracy, and results with and without this correction were compared. A radon-progeny discrimination method was also used to assess its impact on manned-helicopter measurements. Furthermore, development of unmanned airplane monitoring technologies was advanced to enhance the efficiency of wide-area surveys.

Demonstration of the inherent safety feature of HTGRs through the loss-of-forced-cooling test in the HTTR

Nagasumi, Satoru; Hasegawa, Toshinari; Iigaki, Kazuhiko; Nakagawa, Shigeaki; Kubo, Shinji; Shimazaki, Yosuke; Nakajima, Kunihiro; Sakurai, Yosuke; Shinohara, Masanori; Saito, Kenji; et al.

Nuclear Engineering and Design, 446(Part A), p.114542_1 - 114542_14, 2026/01

https://doi.org/10.1016/j.nucengdes.2025.114542

To demonstrate HTGR's safety features, a loss-of-forced-cooling (LOFC) test was conducted using the HTTR. In this test, the forced cooling in the reactor core was intentionally lost by shutting down all helium gas circulators (HGCs) without reactor scram. During steady-state operation at 100% reactor power (30 MW), after the LOFC, the reactor power spontaneously decreased. This power reduction occurred due to the negative reactivity feedback effect triggered by an increase in core temperature. The power stabilized at a lower value of 1.2% after re-criticality. Additionally, the measured radioactivity concentration in the primary coolant remained nearly unchanged during this LOFC operation and during an immediately subsequent HTTR operation. This indicates no failure of the coated particle fuel, even after the increase in core temperature associated with the LOFC event. These results provide experimental evidence of the safety features of HTGRs.

Achievement of safety demonstration tests using HTTR; Loss of forced cooling test at 100% reactor power (30 MW)

Nagasumi, Satoru; Hasegawa, Toshinari; Nakagawa, Shigeaki; Kubo, Shinji; Iigaki, Kazuhiko; Shinohara, Masanori; Saikusa, Akio; Nojiri, Naoki; Saito, Kenji; Furusawa, Takayuki; et al.

JAEA-Research 2025-005, 23 Pages, 2025/07

JAEA-Research-2025-005.pdf:2.68MB

A safety demonstration test under abnormal operating conditions using the HTTR (High Temperature Engineering Test Reactor) was conducted to demonstrate safety features of the HTGRs (High Temperature Gas-cooled Reactors). Under a simulation of a control rod shutdown failure, all primary helium gas circulators were intentionally stopped during a steady-state operation at 100% reactor thermal power (30 MW), temporal changes of the reactor power and temperatures around the reactor pressure vessel (RPV) were obtained after the complete loss of forced heat removal from the reactor core. After the event (primary coolant flow stopped), the reactor power quickly decreased due to the negative reactivity feedback associated with the core temperature rise, and then the reactor power spontaneously shifted to a stable state of low power (about 1.2%) even after a recriticality. Heat dissipation from RPV surface to a surrounding vessel cooling system (water-cooled panels) ensured the amount of heat removal required to maintain the reactor temperature constant in the low power state. In this way, the transition from the event occurrence to the stable and safety state, i.e., inherent safety features of HTGRs, were demonstrated in the case of core forced cooling loss without active shutdown operations.

Development of a bio-based adhesive by polymerization of Boc-protected vinyl catechol derived from caffeic acid

Tanizaki, Shiho*; Kubo, Tomohiro*; Bito, Yosuke*; Mori, Shigeki*; Aoki, Hiroyuki; Sato, Kotaro*

RSC Sustainability (Internet), 3(4), p.1714 - 1720, 2025/04

https://doi.org/10.1039/D4SU00629A

Gyro-spintronic material science using vorticity gradient in solids

Nozaki, Yukio*; Sukegawa, Hiroaki*; Watanabe, Shinichi*; Yunoki, Seiji*; Horaguchi, Taisuke*; Nakayama, Hayato*; Yamanoi, Kazuto*; Wen, Z.*; He, C.*; Song, J.*; et al.

Science and Technology of Advanced Materials, 26(1), p.2428153_1 - 2428153_39, 2025/02

https://doi.org/10.1080/14686996.2024.2428153

Quantitative evaluation of leakage flow rate in the sealing part using graphite gland packing to mount a hydrogen separation membrane tube for HI decomposition membrane reaction

Sugimoto, Chihiro; Myagmarjav, O.; Tanaka, Nobuyuki; Noguchi, Hiroki; Takegami, Hiroaki; Kubo, Shinji

International Journal of Hydrogen Energy, 95, p.98 - 107, 2024/12

https://doi.org/10.1016/j.ijhydene.2024.10.334

Progress of sodium-cooled fast reactor developments in Japan taking into account total lifecycle, risk-informed approach, and sustainability

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

https://doi.org/10.1016/j.nucengdes.2024.113062

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.

Radiation monitoring via manned helicopter around the nuclear power station in the fiscal year 2022 (Contract research)

Futemma, Akira; Sanada, Yukihisa; Nagakubo, Azusa; Kawasaki, Yoshiharu*; Iwai, Takeyuki*; Hiraga, Shogo*; Haginoya, Masashi*; Matsunaga, Yuki*; Akutsu, Yuichiro*; Arai, Yoshinori*; et al.

JAEA-Technology 2023-027, 146 Pages, 2024/03

JAEA-Technology-2023-027.pdf:18.12MB

By the accident at Tokyo Electric Power Company's (TEPCO's) Fukushima Daiichi Nuclear Power Station (FDNPS), caused by tsunami triggered by the 2011 off the Pacific coast of Tohoku Earthquake, a large amount of radioactive material was released into the surrounding environment. After the accident, Airborne Radiation Monitoring (ARM) via manned helicopter has been applied as a method to quickly and extensively measure the distribution of radiation. Japan Atomic Energy Agency (JAEA) has continuously conducted ARM via manned helicopter around FDNPS. In this report, we summarize the results of the ARM around FDNPS in the fiscal year 2022, evaluate the changes of ambient dose rates and other parameters based on the comparison to the past ARM results, and discuss the causes of such changes. In order to contribute to improve the accuracy of ambient dose rate conversion, we analyzed the ARM data taking into account undulating topography, and evaluated the effect of this method. Furthermore, the effect of radon progenies in the air on the ARM was evaluated by applying the discrimination method to the measurement results.

Simulation-based dynamic probabilistic risk assessment of an internal flooding-initiated accident in nuclear power plant using THALES2 and RAPID

Kubo, Kotaro; Zheng, X.; Tanaka, Yoichi; Tamaki, Hitoshi; Sugiyama, Tomoyuki; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*

Proceedings of the Institution of Mechanical Engineers, Part O; Journal of Risk and Reliability, 237(5), p.947 - 957, 2023/10

https://doi.org/10.1177/1748006X221091604

Probabilistic risk assessment (PRA) is a method used to assess the risks associated with large and complex systems. However, the timing at which nuclear power plant structures, systems, and components are damaged is difficult to estimate if the risk of an external event is evaluated using conventional PRA based on event trees and fault trees. A methodology coupling thermal-hydraulic analysis with external event simulations using Risk Assessment with Plant Interactive Dynamics (RAPID) is therefore proposed to overcome this limitation. A flood propagation model based on Bernoulli's theorem was applied to represent internal flooding in the turbine building of the pressurized water reactor. Uncertainties were also taken into account, including the flow rate of the floodwater source and the failure criteria for the mitigation systems. The simulated recovery actions included the operator isolating the floodwater source and using a drainage pump; these actions were modeled using several simplifications. Overall, the results indicate that combining isolation and drainage can reduce the conditional core damage probability upon the occurrence of flooding by approximately 90%.