International benchmark on radionuclide solubility and speciation for the Long-term In-situ Test (LIT) at Grimsel Test Site (GTS, Switzerland)

Noseck, U.*; Schfer, T.*; Alonso, U.*; Hamamoto, Takafumi*; Havlova, V.*; Hibberd, R.*; Ishidera, Takamitsu; Kitamura, Akira; Klajmon, M.*; Missana, T.*; et al.

Applied Geochemistry, 201, p.106762_1 - 106762_23, 2026/04

https://doi.org/10.1016/j.apgeochem.2026.106762

Thermodynamic benchmark calculations have been performed to better understand the behavior of Se(VI), Tc(VII), U(VI), Np(V), Am(III), Th(IV) and Pu(IV)) in the evolving geochemical conditions of the Long-term In-situ Test (LIT) at the Grimsel Test Site (GTS) and corresponding mock-up experiment. It also aims to identify the status of the geochemical speciation models and databases for these elements. The experiments are simulating the near-field conditions in some radioactive waste repository concept including a bentonite engineered barrier emplaced in crystalline rock and the findings are contributing to the long-term safety assessment of these facilities.

OSCAAR version 2.0 User's Manual

Risk Analysis Research Group, Nuclear Safety Research Center

JAEA-Testing 2025-007, 110 Pages, 2026/03

JAEA-Testing-2025-007.pdf:2.57MB

The Japan Atomic Energy Agency's Nuclear Safety Research Center is developing the Level 3 PRA code OSCAAR as part of its research on probabilistic risk assessment (PRA) for nuclear power plant accidents. OSCAAR is a computational code that evaluates the advection, diffusion, and deposition of radioactive materials released into the environment under various meteorological conditions, based on source terms obtained from Level 2 PRA. It can probabilistically assess the radiation doses and health effects to the public caused by these radioactive materials. OSCAAR can account for the dose reduction effects of protective measures implemented during an actual nuclear power plant accident, thereby contributing to the pre-planning of countermeasures and plans to reduce the exposure of residents near nuclear power plants during an accident. This report is a manual for users to create input files and execute the OSCAAR program.

Report on research and development of neutron beam utilization at JRR-3 for internal use in FY2023 and FY2024

Materials Sciences Research Center

JAEA-Review 2025-058, 175 Pages, 2026/03

JAEA-Review-2025-058.pdf:8.17MB

Fifteen neutron beam experimental instruments managed by JAEA are installed in JRR-3 (Japan Research Reactor No.3) and are available for internal use including upgrading of instruments and for external users to produce various research results. This report summarizes the progress of internal application research and technical development such as upgrading of neutron beam instruments in the fiscal years 2023 and 2024 after the restart of operation.

Synthesis of high-entropy hydride from the cantor alloy (fcc-CoCrFeNiMn) at extreme conditions

Glazyrin, K.*; Spektor, K.*; Bykov, M.*; Carvalho, P. H. B.*; Dong, W.*; Krmann, F.*; Sano, Asami; Hattori, Takanori; Beyer, D. C.*; Sahlberg, M.*; et al.

Nature Communications (Internet), 17, p.2622_1 - 2622_10, 2026/03

https://doi.org/10.1038/s41467-026-70483-3

Studies of high entropy materials contribute to various fields of science and reveal ever more exciting properties of applied interest. Here, we perform a study of the resistance of a Cantor alloy (CoCrFeNiMn) to hydrogen through high-pressure experiments at elevated temperatures by X-ray and neutron time-of-flight experiments and ab initio calculations. We report formation of an fcc hydride based on the Cantor alloy composition. We also provide its characterization, including an estimate of hydrogen content. These findings contribute to the growing body of knowledge on the complex chemistry of high-entropy alloys and high-entropy-hydrides.

Orbital angular momentum of azimuthal spin waves

Valet, T.*; Yamamoto, Kei; Pigeau, B.*; de Loubens, G.*; Klein, O.*

Physical Review B, 113, p.L100410_1 - L100410_6, 2026/03

https://doi.org/10.1103/l1x7-t8pn

Model descriptions for OSCAAR version 2.0

Risk Analysis Research Group, Nuclear Safety Research Center

JAEA-Data/Code 2025-015, 68 Pages, 2026/02

JAEA-Data-Code-2025-015.pdf:2.82MB

The Japan Atomic Energy Agency's Nuclear Safety Research Center is developing the Level 3 PRA code OSCAAR as part of its research on probabilistic risk assessment (PRA) for nuclear power plant accidents. OSCAAR is a computational code that evaluates the advection, diffusion, and deposition of radioactive materials released into the environment under various meteorological conditions, based on source terms obtained from Level 2 PRA. It can probabilistically assess the radiation doses and health effects to the public caused by these radioactive materials. OSCAAR can account for the dose reduction effects of protective measures implemented during an actual nuclear power plant accident, thereby contributing to the preplanning of countermeasures and plans to reduce the exposure of residents near nuclear power plants during an accident. This report is a manual explaining the analysis model used in OSCAAR code version 2.0.

New insights into the Oklo natural fission reactors elucidated from isotopic studies using state-of-the-art analytical techniques

Hidaka, Hiroshi*; Kagami, Saya; Saito, Takaharu; Yokoyama, Tatsunori; Pravdivtseva, O.*; Meshik, A.*; Bentridi, S.*; Durastanti Rabnga Mombo, E. D.*; Gall, B.*

Radiation Protection Dosimetry, 202(2), p.65 - 69, 2026/02

https://doi.org/10.1093/rpd/ncaf119

Isotopic studies of the Oklo natural fission reactors have provided very important information to characterize the operating conditions of the individual reactor zones (RZs). Most of the isotopic data of the Oklo RZ materials were given more than a quarter of a century ago. Significant progress has recently been made in the instrumental developments of mass spectrometry for isotopic analyses. If more precise isotopic data from the RZ materials were obtained, we may gain new insights into understanding the Oklo phenomenon. There is still much to learn from the Oklo RZs. In this paper, we introduce our new approaches to elucidate the Oklo phenomenon from isotopic studies using state-of-the-art analytical techniques.

Neutron total and capture cross-section measurement and resolved resonance analysis of Er

Rovira Leveroni, G.; Kimura, Atsushi; Nakamura, Shoji; Endo, Shunsuke; Iwamoto, Osamu; Iwamoto, Nobuyuki; Katabuchi, Tatsuya*

Annals of Nuclear Energy, 225, p.111688_1 - 111688_18, 2026/01

https://doi.org/10.1016/j.anucene.2025.111688

Possible transition from silandic to aluandic andosols with acidification as evidenced by detailed profile investigation of two well-preserved forest andosols in Japan

Alam, M. M.*; Yamakita, Eri*; Inoue, Yuzuru*; Koarashi, Jun; Atarashi-Andoh, Mariko; Abe, Yukiko; Nakayama, Masataka*; Mori, Yuki*; Hiradate, Shuntaro*

Soil Science and Plant Nutrition, 72(1), p.55 - 66, 2026/01

https://doi.org/10.1080/00380768.2025.2511803

Pilot study on thermal, physico-chemical, and mechanical behavior of concrete to understand the failure behavior of Fukushima Daiichi Nuclear Power Station reactor pressure vessel pedestals (Contract research); FY2023 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; Tokai National Higher Education and Research System*

JAEA-Review 2025-034, 83 Pages, 2025/12

JAEA-Review-2025-034.pdf:6.9MB

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 (1F), 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 FY2023, this report summarizes the research results of the "Pilot study on thermal, physico-chemical, and mechanical behavior of concrete to understand the failure behavior of Fukushima Daiichi Nuclear Power Station reactor pressure vessel pedestals" conducted in FY2023. The present study aims to examine the mechanism of the collapse of only concrete with rebar remaining in the pedestal in the containment vessel (PCV) of 1F. In verifying concrete-specific factors, (1) to clarify the short-term dissolution mechanism by high temperature, we investigated data acquisition methods in melting experiments, established an analytical framework for determining dissolution, and developed a numerical analysis method for volume change by heating. Additionally, (2) to clarify long-term dissolution mechanism by temperature history, we organized the temperature and water injection history, determined concrete exposure conditions during experiments, and established a method for selecting materials and measuring expansion. Furthermore, we summarized existing knowledge of the expansion phenomenon caused by water supply after high temperature heating. In the verification of special external environmental factors, (1) to evaluate thermal conditions of PCV concrete during an accident, a preliminary heat transfer analysis of fuel debris was conducted. In addition, (2) as elemental behavior tests and comprehensive tests, a preliminary high temperature storage test on concrete materials in a water vapor atmosphere and a preliminary reaction test on the reaction behavior of metal debris and concrete were conducted. Furthermore, uranium-containing suboxides were prepared. This study provided comprehensive insight into the mechanism of concrete failure in 1F Unit 1.