Densities, surface tensions, and viscosities of molten high-silicon electrical steels with different silicon contents

Neubert, L.*; Bell, M. R.*; Yamamoto, Taisei*; Nishi, Tsuyoshi*; Yamano, Hidemasa; Ahrenhold, F.*; Volkova, O.*

Steel Research International, 96(5), p.202400237_1 - 202400237_8, 2025/05

https://doi.org/10.1002/srin.202400237

Density of a molten stainless steel-BC alloy measured in the electrostatic levitation furnace onboard the international space station

Ishikawa, Takehiko*; Oda, Hirohisa*; Koyama, Chihiro*; Shimonishi, Rina*; Ikeuchi, Rumiko*; Paradis, P.-F.*; Okada, Jumpei*; Fukuyama, Hiroyuki*; Yamano, Hidemasa

International Journal of Microgravity Science and Application, 42(2), p.420202_1 - 420202_10, 2025/04

https://doi.org/10.15011/jasma.42.420202

Density, surface tension, and viscosity of liquid low-sulfur manganese-boron steel via maximum bubble pressure and oscillating crucible methods

Bell, M. R.*; Neubert, L.*; Sherstneva, A.*; Yamamoto, Taisei*; Nishi, Tsuyoshi*; Yamano, Hidemasa; Weinberg, M.*; Volkova, O.*

Steel Research International, p.2400252_1 - 2400252_10, 2025/00

https://doi.org/10.1002/srin.202400252

In this study, the thermophysical properties of low-sulfur manganese-boron steel with varying boron and sulfur contents at different temperatures are investigated.

A Preliminary study for boron mixing effect on severe accident scenario in sodium-cooled fast reactor

Yamano, Hidemasa; Morita, Koji*

Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (Internet), 9 Pages, 2024/11

SIMMER-III and SIMMER-IV; Computer codes for LMFR core disruptive accident analysis

Kondo, Satoru; Tobita, Yoshiharu*; Morita, Koji*; Kamiyama, Kenji; Yamano, Hidemasa; Suzuki, Toru*; Tagami, Hirotaka; Sogabe, Joji; Ishida, Shinya

JAEA-Research 2024-008, 235 Pages, 2024/10

JAEA-Research-2024-008.pdf:4.77MB

The SIMMER-III and SIMMER-IV computer codes, developed at the Japan Atomic Energy Agency are the codes with two- and three-dimensional, multi-field, multi-component fluid-dynamics models, coupled with a space- and time-dependent neutron kinetics model. The codes have been used widely for simulating complex phenomena during core-disruptive accidents in liquid-metal fast reactors. Advanced features of the codes in comparison with the former codes include: stable and robust fluid-dynamics algorithm with up to 8 velocity fields, improved representation of structures and multi-phase flow topology, comprehensive treatment of complex heat and mass transfer processes, accurate analytic equations of state, a stable and efficient neutron flux shape solution method and decay heat model. This report describes the models and methods of SIMMER-III and SIMMER-IV. For those individual models, the details of which have been reported elsewhere, only the outlines of the models are presented. The reports of code verification and validation have been already published.

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

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

Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10

To enhance the accuracy of the safety evaluations in sodium-cooled fast reactors, it is necessary to develop a method to realistically evaluate the reactivity caused by core deformation. In this regard, Japan and the United States jointly conducted a benchmark analysis of thermal bowing experiments of a single duct of Joyo-type fuel assembly. The aim was to confirm the validity of the core bowing analysis codes. Comparisons of analysis and test results revealed that the core bowing analysis codes used by both countries were able to reasonably predict the axial distribution of horizontal duct displacement of a single duct due to thermal bowing and the contact load on the duct pad.

France-Japan collaboration on severe accident studies in sodium-cooled fast reactors, 3; Thermodynamic, Kinetic, and Thermophysical Studies of Core Material Mixture

Yamano, Hidemasa; Emura, Yuki; Takai, Toshihide; Kubo, Shigenobu; Quaini, A.*; Fossati, P.*; Delacroix, J.*; Journeau, C.*

Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10

This report mainly introduces trends in fast reactor development in Japan in addition to introducing overseas development trends for major developing countries. The paper describes major severe accident study results focusing on kinetics of interaction in core material mixtures, physical properties of core material mixtures, high temperature thermodynamic data for the uranium oxide (UO)-iron (Fe)-boron carbide (BC) system, experimental studies on BC-stainless steel (SS) kinetics and BC-SS eutectic material relocation (freezing), and BC-SS eutectic and kinetics models for severe accident code systems,

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

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

Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10

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

Effectiveness evaluation of the measures for improving resilience at ultra-high temperatures

Onoda, Yuichi; Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa

Proceedings of Probabilistic Safety Assessment and Management & Asian Symposium on Risk Assessment and Management (PSAM17 & ASRAM2024) (Internet), 10 Pages, 2024/10

We developed the measures for improving resilience of the sodium-cooled fast reactor structure using the failure mitigation technology and evaluated the effectiveness of the measures. To prevent core damage in the event of an accident progressing to an ultra-high temperature state, both measures to prevent overpressure in the reactor vessel and measures to cool the reactor core are required. As a core cooling measure, we developed a core cooling concept that promotes radiant heat transfer from the reactor vessel and cools the containment vessel outer surface by natural convection named Containment Vessel Auxiliary Cooling System (CVACS). We developed a method to use the reduction rate of core damage frequency as an indicator for effectiveness of the measures for improving resilience. The core damage frequency was evaluated by calculating the core cooling performance using CVACS, reflecting the results of structural analysis and human reliability analysis. By implementing measures for improving resilience in addition to existing measures, the core damage frequency of Japan loop-type sodium-cooled fast reactor caused by LOHRS has been reduced to about one-hundredth of the previous level.

Effectiveness evaluation of the measures for improving resilience of nuclear structures against excessive earthquake, 1; Fragility evaluation of reactor vessel based on structural analysis

Nishino, Hiroyuki; Kurisaka, Kenichi; Futagami, Satoshi; Watakabe, Tomoyoshi; Yamano, Hidemasa

Proceedings of Probabilistic Safety Assessment and Management & Asian Symposium on Risk Assessment and Management (PSAM17 & ASRAM2024) (Internet), 10 Pages, 2024/10

The reactor vessel (RV) buckling was a dominant contributor to core damage. However, even if the RV is buckled due to seismic shaking, it is expected that the RV maintains stable state without unstable failure such as rupture, collapse. Realistic consideration of the post-buckling behavior is regarded as a measure for improving the resilience in this study. The purpose of this study is to understand the post-buckling deformation behavior of the RV and to evaluate the RV fragility based on fatigue failure. This study performed structural analysis using a finite element method to quantify time histories of displacement, strain, etc. As the result of the analysis, wrinkles of the buckling appeared at the elevation higher than the liquid level in the RV. The largest strain value was also indicated around this elevation. The cumulative fatigue damage fraction was evaluated in this analysis to evaluate the fragility of fatigue failure in addition to the buckling fragility. The result showed that the seismic intensity for the median fragility of the fatigue failure was about six times larger than the design-basis ground motion. This is 1.2 times larger than the buckling-based result, which suggests that realistic evaluation of the post-buckling behavior could contribute to improving the resilience of the nuclear structure.