Particle arrangements and optical changes induced by the water swelling of melanin-like polydopamine layers

Watanabe, Taku*; Maejima, Yui*; Ueda, Yuki; Motokawa, Ryuhei; Takabatake, Ai*; Takeda, Shinichi*; Fudoji, Hiroshi*; Kishikawa, Keiki*; Koori, Michinari*

Langmuir, 41(34), p.22762 - 22773, 2025/09

https://doi.org/10.1021/acs.langmuir.5c01855

The assembled structures of melanin particles, i.e., colloidal particles coated with a melanin-like polydopamine (PDA) layer, create vivid structural colors. While the thickness of the PDA layer influences the particle arrangement and optical properties, the underlying mechanism has remained controversial. We demonstrate that the waterswelling characteristics of PDA are crucial factors governing the dispersion and aggregation of these particles in solution. Detailed comparisons between dry and wet conditions revealed that the PDA layer readily absorbs water molecules, which leads to significant swelling in the thicker layers. The swelling of the PDA layers determined whether the particles remained dispersed or partially aggregated in the water, ultimately controlling the particle arrangement in the dry state once the water evaporated. These findings provide insights into the self-assembly of colloidal particles and offer a strategy for tuning the periodic particle order. This feature is pivotal for various applications in optical and sensing technologies.

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.

Non-condensable gas accumulation and distribution due to condensation in the CIGMA Facility; Implications for Fukushima Daiichi Unit 3 (1F3)

Hamdani, A.; Soma, Shu; Abe, Satoshi; Shibamoto, Yasuteru

Progress in Nuclear Energy, 185, p.105771_1 - 105771_13, 2025/07

https://doi.org/10.1016/j.pnucene.2025.105771

Human resource development related to remote control technology for monitoring inside RPV pedestal during retrieval of fuel debris (Contract research); FY2023 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; The University of Tokyo*

JAEA-Review 2025-001, 94 Pages, 2025/06

JAEA-Review-2025-001.pdf:6.21MB

The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2023. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. (TEPCO). For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2019, this report summarizes the research results of the "Human resource development related to remote control technology for monitoring inside RPV pedestal during retrieval of fuel debris" conducted from FY2019 to FY2023. The present study aims to construct a monitoring platform for understanding the status inside a reactor during fuel debris removal, and measurement and visualization by sensors moving on the platform. In addition, to develop research personnel through research education by participating in such research projects, classroom lectures, and facility tours is also a goal of this project. In FY2023, along with the verification of each system, a three-dimensional reconstruction model was generated using images acquired from a moving camera on the monitoring platform in a simulated environment, and an integrated experiment was conducted to demonstrate that it is possible to present images from the optimal viewpoint for the visualization target, with the cooperation of each research theme.

Detailed computational models for nuclear criticality analyses on the first startup cores of NSRR: A TRIGA annular core pulse reactor

Yanagisawa, Hiroshi; Motome, Yuiko

JAEA-Research 2025-001, 99 Pages, 2025/06

JAEA-Research-2025-001.pdf:1.98MB

The detailed computational models for nuclear criticality analyses on the first startup cores of NSRR (Nuclear Safety Research Reactor), which is categorized as a TRIGA-ACPR (Annular Core Pulse Reactor), were created for the purposes of deeper understandings of safety inspection data on the neutron absorber rod worths of reactivity and improvement of determination technique of the reactivity worths. The uncertainties in effective neutron multiplication factor (k) propagated from errors in the geometry, material, and operation data for the present models were evaluated in detail by using the MVP version 3 code with the latest Japanese nuclear data library, JENDL-5, and the previous versions of JENDL libraries. As a result, the overall uncertainties in k for the present models were evaluated to be in the range of 0.0027 to 0.0029 k. It is expected that the present models will be utilized as the benchmark on k for TRIGA-ACPR. Moreover, it is confirmed that the overall uncertainties were sufficiently smaller than the values of absorber rod worths determined in NSRR. Thus, it is also considered that the present models are applicable to further analyses on the absorber rod worths in NSRR.

Numerical simulation of coupled THM behaviour of full-scale EBS in backfilled experimental gallery in the Horonobe URL

Sugita, Yutaka; Ono, Hirokazu; Beese, S.*; Pan, P.*; Kim, M.*; Lee, C.*; Jove-Colon, C.*; Lopez, C. M.*; Liang, S.-Y.*

Geomechanics for Energy and the Environment, 42, p.100668_1 - 100668_21, 2025/06

https://doi.org/10.1016/j.gete.2025.100668

The international cooperative project DECOVALEX 2023 focused on the Horonobe EBS experiment in the Task D, which was undertaken to study, using numerical analyses, the thermo-hydro-mechanical (or thermo-hydro) interactions in bentonite based engineered barriers. One full-scale in-situ experiment and four laboratory experiments, largely complementary, were selected for modelling. The Horonobe EBS experiment is a temperature-controlled non-isothermal experiment combined with artificial groundwater injection. The Horonobe EBS experiment consists of the heating and cooling phases. Six research teams performed the THM or TH (depended on research team approach) numerical analyses using a variety of computer codes, formulations and constitutive laws.

Pressure-induced elongation of hydrogen-oxygen bond in sodium silicate melts

Ohashi, Tomonori*; Sakamaki, Tatsuya*; Funakoshi, Kenichi*; Steinle-Neumann, G.*; Hattori, Takanori; Yuan, L.*; Suzuki, Akio*

Journal of Mineralogical and Petrological Sciences (Internet), 120(1), p.240926a_1 - 240926a_13, 2025/06

https://doi.org/10.2465/jmps.240926a

We explore the structures of dry and hydrated (HO and DO) NaSiO melt at 0-6 GPa and 1000-1300 K and glasses recovered from high pressure and temperatures by in-situ neutron and X-ray diffraction. The structures of the melts at 0-10 GPa and 3000 K are also investigated by ab-initio molecular dynamics simulation. In-situ neutron experiments revealed that the D-O distance increases with compression due to the formation of -O-D-O- bridging species, which is reproduced by the molecular dynamics simulations. The pressure-induced -O-D-O- formation reflects a more rigid incorporation of hydrogen, which acts as a mechanism for the experimentally observed higher solubility of water in silicate melts. Together with shrinking modifier domains, this process dominates the compression behavior of hydrous NaSiO melt, whereas the compression of dry NaSiO at 0-10 GPa and 3000 K is governed largely by bending of the Si-O-Si angle. The molecular dynamics simulations on hydrous NaSiO melts further suggest that the sodium ions are scavenged from its network-modifying role via 2(Si-O + Na) Si-(O-Si-O) + 2Na and Si-O + Na + Si-OH Si-(O-H-O-Si) + Na with increasing pressure.

Experiments on central reaction rate ratios and fission distributions in the FCA-XXII-1 assembly simulating highly enriched MOX fueled tight lattice LWR cores

Fukushima, Masahiro; Ando, Masaki; Nagaya, Yasunobu

Nuclear Science and Engineering, 199(6), p.1029 - 1043, 2025/06

https://doi.org/10.1080/00295639.2024.2405668

Numerical analysis of a potential Reactor Pressure Vessel (RPV) boundary failure mechanism in Fukushima Daiichi Nuclear Power Station Unit-2

Li, X.; Yamaji, Akifumi*; Sato, Ikken*; Yamashita, Takuya

Annals of Nuclear Energy, 214, p.111217_1 - 111217_13, 2025/05

https://doi.org/10.1016/j.anucene.2025.111217

Development of corrosion-stable dual-Si-layered membranes for hydrogen production via thermochemical iodine-sulfur process

Myagmarjav, O.; Tanaka, Nobuyuki; Noguchi, Hiroki; Kamiji, Yu; Ono, Masato; Nomura, Mikihiro*; Takegami, Hiroaki

Progress in Nuclear Science and Technology (Internet), 7, p.235 - 242, 2025/05

https://doi.org/10.15669/pnst.7.235