Hydrogen in the Earth core inferred from neutron imaging and diffraction

Takahashi, Naoki*; Sakamaki, Tatsuya*; Hattori, Takanori; Funakoshi, Kenichi*; Arima-Osonoi, Hiroshi*; Sano, Asami; Abe, Jun*; Suzuki, Akio*

Scientific Reports (Internet), 16, p.14162_1 - 14162_13, 2026/05

https://doi.org/10.1038/s41598-026-49969-z

We performed high-pressure and high-temperature neutron diffraction and imaging experiments in situ to determine the hydrogen content in liquid iron. We observed that liquid iron contains 0.17(3) wt.% H at 3.4 GPa and 1400 K, indicating that liquid iron is hydrogenated in the magma ocean during core formation. For the hydrogen content in the liquid iron at the base of the magma ocean, we estimated that the outer and inner cores contain 0.60-0.72 and 0.30-0.44 wt.% H, corresponding to 70-85 and 1.9-2.7 times the mass of hydrogen in the ocean, respectively. This suggests that hydrogen can contribute more than half of the density deficit in the outer core. For the magma ocean equilibrating with the hydrogen-rich primary atmosphere, the study findings show that liquid iron plays a crucial role in transporting a large amount of hydrogen into the core.

Mechanical property evaluation of rocks around Tanakura Fault, 1; Applicability of impact and needle penetration tests

Asahina, Daisuke*; Takamura, Hiroaki*; Tsukamoto, Koji*; Honda, Fumito*; ITO, Takashi*; IWATA, Naoki*; Aydan, mer*; Sato, Toshinori

Dai-51-Kai Gamban Rikigaku Ni Kansuru Shimpojiumu Koen Rombunshu(Internet), p.43 - 47, 2025/12

Fault zones commonly contain distinctive structures such as fault cores and damage zones, where physical properties vary locally. These zones are important for evaluating underground safety and groundwater flow in waste disposal, yet quantitative measurements near faults remain limited. We applied Schmidt hammer and needle penetration tests to assess the mechanical properties of the rocks near the Tanakura Fault, in a simple, non-destructive manner. The results showed a clear spatial variation in measured values with distance from the fault gouge boundary, suggesting that combining methods with different scales and sensitivities can provide a more comprehensive evaluation of rock mass properties near faults.

Re-evaluation of nuclear criticality characteristics for infinite and finite heterogeneous lattice systems composed of uranium-zirconium hydride fuel rods used in the TRIGA annular core pulse reactor NSRR

Yanagisawa, Hiroshi; Motome, Yuiko

JAEA-Research 2025-010, 197 Pages, 2025/11

JAEA-Research-2025-010.pdf:3.5MB

For understandings of nuclear criticality risks of TRIGA fuel rods and review of safety measures for handling them, nuclear criticality characteristics for infinite and finite heterogeneous lattice systems composed of the NSRR fuel rods were re-evaluated with the use of a detailed computational model for the fuel rod. The MVP version 3 code was used with the JENDL libraries including the latest version, JENDL-5, for the re-evaluation. As the criticality characteristics, variations of neutron multiplication factors of the infinite and water-reflected finite systems were examined in detail with parameters of the lattice pitch and density of moderator water. From the results of the re-evaluated criticality characteristics, the minimum critical number of fuel rods for the water-reflected hexagonal shaped lattice system was obtained to be 46.8 0.2 using the JENDL-5 library. Moreover, the attainability of criticality without the water as moderator and reflector was examined because the zirconium hydride moderator and graphite reflector are equipped with the TRIGA fuel rod. It was found that the criticality is possible to be attained by 115.7 0.6 of the number of fuel rods, which is the smaller number of fuel rods than loaded in the NSRR standard core, even though no water exists.

Preliminary study of diffusion and SP3 calculations in unstructured mesh geometry for core deformation reactivity evaluation on SFR

Kato, Shinya; Doda, Norihiro; Yokoyama, Kenji; Tanaka, Masaaki; Endo, Tomohiro*

Proceedings of 2025 International Congress on Advances in Nuclear Power Plants (ICAPP 2025) (Internet), 11 Pages, 2025/09

During a reactor power increase in ULOF and UTOP events in sodium-cooled fast reactors, core deformation due to thermal expansion of core elements is expected to cause a negative feed-back effect to suppress this power increase. An analytical evaluation method of core deformation reactivity for design is being developed in JAEA. However, the neutronics calculation module uses several approximations. This study aims to develop the detailed evaluation method as a reference neutron transport calculation code for confirmation of the validity of the calculated core deformation reactivity. Here, the two-dimensional finite volume method (FVM) code based on simplified P3 (SP3) approximation with unstructured mesh have been developed as the first step of the development. This paper describes the calculation theory of the FVM code, the procedure of introducing SP3 approximation into the code and the verification results of the functions developed.

Validation study on SFR core bowing codes using Joyo ex-core experiment data; Multiple duct bowing benchmark

Wozniak, N.*; Ohgama, Kazuya; Doda, Norihiro; Ota, Hirokazu*; Shemon, E.*; Feng, B.*; Uwaba, Tomoyuki; Futagami, Satoshi; Tanaka, Masaaki; Yamano, Hidemasa; et al.

Progress in Nuclear Science and Technology (Internet), 8, p.165 - 170, 2025/09

https://doi.org/10.15669/pnst.8.165

To enhance the accuracy of the safety evaluations for sodium-cooled fast reactors, it is necessary to develop a method that can realistically evaluate the reactivity changes induced by core deformation. In this context, Japan and the United States jointly conducted a benchmark analysis of thermal bowing experiments using multiple ducts from a Joyo-type fuel assembly. The aim was to confirm the validity of the core bowing analysis codes. Comparisons of analysis and experimental results demonstrated that the codes used by both countries were able to reasonably predict the thermal bowing of a row of assemblies in multiple duct configuration.

Analysis and quantification of tritium and other low-level radionuclides present in ALPS-treated water

Shimada, Asako; Hemmi, Ko; Ohira, Saki; Iida, Yoshihisa

Analytical Sciences, 41, p.1383 - 1391, 2025/06

https://doi.org/10.1007/s44211-025-00803-y

Development of gas entrainment evaluation model based on distribution of pressure along vortex center line; Application to a gas entrainment experiment with traveling vortices in an open water channel flow?

Matsushita, Kentaro; Ezure, Toshiki; Tanaka, Masaaki; Imai, Yasutomo*; Fujisaki, Tatsuya*; Sakai, Takaaki*

Nuclear Engineering and Design, 432, p.113785_1 - 113785_16, 2025/02

https://doi.org/10.1016/j.nucengdes.2024.113785

Establishing an evaluation method for the gas entrainment (GE) of argon cover gas due to surface vortices is required in terms of safety design of sodium-cooled fast reactors. To modify the evaluation model in an in-house evaluation tool for GE, StreamViewer, a modified evaluation model on the pressure distribution along the vortex center line (PVL model) was proposed to identify the vortex center lines by connecting continuous vortex center points from the suction port to the surface and evaluate gas core length based on the balance between the hydrostatic pressure and the pressure decrease distribution along the vortex center line. PVL model was applied the three-dimensional numerical analysis results for the experiments where a plate induced unsteady traveling vortices in the open channel flow. Consequently, the GE evaluation using StreamViewer with PVL model could reproduce the relation between the inlet flow velocity and the gas core length in the unsteady vortex flow experiments.

Data report of ROSA/LSTF experiment SB-PV-03; 0.2% pressure vessel bottom break LOCA with SG depressurization and gas inflow

Takeda, Takeshi

JAEA-Data/Code 2024-014, 76 Pages, 2024/12

JAEA-Data-Code-2024-014.pdf:4.0MB

An experiment denoted as SB-PV-03 was conducted on November 19, 2002 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment SB-PV-03 simulated a 0.2% pressure vessel bottom small-break loss-of-coolant accident in a pressurized water reactor (PWR). The test assumptions included total failure of high pressure injection system of emergency core cooling system (ECCS) and noncondensable gas (nitrogen gas) inflow to the primary system from accumulator (ACC) tanks of ECCS. Secondary-side depressurization of both steam generators (SGs) as an accident management (AM) action to achieve the depressurization rate of 55 K/h in the primary system was initiated 10 min after the generation of a safety injection signal, and continued afterwards. Auxiliary feedwater injection into the secondary-side of both SGs was started for 30 min with some delay after the onset of the AM action. The AM action was effective on the primary depressurization until the ACC tanks began to discharge nitrogen gas into the primary system. The core liquid level recovered in oscillative manner because of intermittent coolant injection from the ACC system into both cold legs. Therefore, the core liquid level remained at a small drop. The pressure difference between the primary and SG secondary sides became larger after nitrogen gas ingress. Core uncovery occurred by core boil-off during reflux condensation in the SG U-tubes under nitrogen gas influx. When the maximum cladding surface temperature of simulated fuel rods exceeded the pre-determined value of 908 K, the core power was automatically reduced to protect the LSTF core. After the automatic core power reduction, coolant injection from low pressure injection (LPI) system of ECCS into both cold legs led to the whole core quench. After the continuous core cooling was confirmed through the actuation of the LPI system, the experiment was terminated.

The Effect of a cyclic bending load on the bending resistance of ballooned, ruptured, and oxidized Zircaloy-4 cladding

Li, F.; Narukawa, Takafumi; Udagawa, Yutaka

Journal of Nuclear Science and Technology, 61(8), p.1036 - 1047, 2024/08

https://doi.org/10.1080/00223131.2023.2293100

Report of summer holiday practical training on 2023

Ishitsuka, Etsuo; Nagasumi, Satoru; Hasegawa, Toshinari; Kawai, Hiromi*; Wakisaka, Shinji*; Nagase, Sota*; Nakamura, Kento*; Yaguchi, Hiroki*; Ishii, Toshiaki; Nakano, Yumi*; et al.

JAEA-Technology 2024-008, 23 Pages, 2024/07

JAEA-Technology-2024-008.pdf:1.69MB

Five people from three universities participated in the 2023 summer holiday practical training with the theme of "Technical development on HTTR". The participants practiced the analysis of HTTR core, the analysis of behavior on loss of forced cooling test, the analysis of Iodine deposition behavior in primary cooling system and the feasibility study of energy storage system for HTGRs. In the questionnaire after this training, there were impressions such as that it was useful as a work experience and some students found it useful for their own research. These impressions suggest that this training was generally evaluated as good.