Development of hydrogen behavior integrated analysis system

Terada, Atsuhiko; Thwe Thwe, A.; Hino, Ryutaro*; Harai, Yasutaka*; Sasaki, Gaku*; Shingeya, Hideshi*; Yamashita, Toshiyuki*; Yoneda, Jiro*; Okabayashi, Kazuki*; Sakamoto, Hiroyuki*; et al.

JAEA-Data/Code 2025-012, 151 Pages, 2025/12

JAEA-Data-Code-2025-012.pdf:9.69MB

Based on the lessons learned from the Fukushima Daiichi Nuclear Power Station accident, we have highly paid attention to the advancement of the fundamental technologies which are indispensable in timely response to hydrogen safety measures and assessments especially in both nuclear reactors and decommissioning. Focusing on this attention, we developed an analysis system that predicts the behavior of hydrogen from generation to diffusion, combustion and explosion. The system utilizes the commercial computational fluid dynamics software (FLUENT, AUTODYN), and incorporates new modules and pre/post-processors in order to withstand the general practical use. We also developed a system by utilizing open-source code (OpenFOAM) that can be used in hydrogen disaster prevention plans for nuclear facilities. So far, we have expanded the system to deal with the phenomena that should be considered from the practical point of view for PWR (Pressurized Water Reactor) in nuclear power plants. This report summarizes the overview of the integrated analysis system for hydrogen behavior, the handling method, and real scale analysis examples.

Core degradation behavior under ULOF transient on CFV-type core equipped SFR

Onoda, Yuichi; Kubota, Ryuzaburo*; Yamano, Hidemasa

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

Steam Explosion Simulation Code JASMINE v.3 User's Guide; Revised for code version 3.3c

Iwasawa, Yuzuru; Matsumoto, Toshinori; Moriyama, Kiyofumi*

JAEA-Data/Code 2025-001, 199 Pages, 2025/06

JAEA-Data-Code-2025-001.pdf:9.71MB

A steam explosion is defined as a phenomenon that occurs when a hot liquid comes into contact with a low-temperature cold liquid with volatile properties. The rapid transfer of heat from the hot liquid to the cold liquid results in a chain reaction of the explosive vaporization of the cold liquid and fine fragmentation of the hot liquid. The explosive vaporization of the cold liquid initiates the propagation of shock waves in the cold liquid. The expansion of the hot and cold liquid mixture exerts mechanical forces on the surrounding structures. In severe accidents of light water reactors, the molten core (melt) is displaced into the coolant water, resulting in fuel-coolant interactions (FCIs). The explosive FCI, referred to as a steam explosion, has been identified as a significant safety assessment issue as it can compromise the integrity of the primary containment vessel. The JASMINE code is an analytical tool developed to evaluate the mechanical forces imposed by steam explosions in nuclear reactors. It performs numerical simulations of steam explosions in a mechanistic manner. The present report describes modeling concepts, basic equations, numerical solutions, and example simulations, as well as instructions for input preparation, code execution, and the use of supporting tools for practical purpose. The present report is the updated version of the "Steam Explosion Simulation Code JASMINE v.3 User's Guide, JAEA-Data/ Code 2008-014". The present report was compiled and updated based on the latest version of the code, JASMINE 3.3c, with corrections for minor errors of the distributed code JASMINE 3.3b, and conformance to recently widely used compilers on UNIX-like environments such as the GNU compiler. The numerical simulations described in the present report were obtained using the latest version JASMINE 3.3c. The latest parameter adjustment method for a model parameter proposed by the previous study is employed to conduct the numerical simulations.

Integrity evaluation of boundary function of main components in nuclear plants during severe accidents

Tsukimori, Kazuyuki; Yada, Hiroki

Journal of Pressure Vessel Technology, 147(3), p.031901_1 - 031901_9, 2025/06

https://doi.org/10.1115/1.4067760

After the accident at the Fukushima Daiichi Nuclear Power Plant, very strict safety measures were implemented for nuclear power plants in Japan. It thus becomes a crucial issue if the safety of a plant is maintained or not at beyond design basis events. In this study, head plates and bellows were examined as components that compose the parts of the boundary of vessels that contain the primary coolant of a prototype fast breeder reactor. The behaviors of buckling, post-buckling deformation, and penetration failure, that is, loss of boundary function of these components with increasing pressure were investigated. The series of this research program started in FY2013 and the research proceeded step by step. The new result in this paper is the application of the proposed criteria to head plates and bellows, and a conservative estimation of penetration failure of these components is obtained.

Handbook of Advanced Nuclear Hydrogen Safety (2nd Edition); Development of hydrogen behavior integrated analysis system and application to actual PWR

Terada, Atsuhiko; Thwe Thwe, A.; Hino, Ryutaro*

JAEA-Review 2024-049, 400 Pages, 2025/03

JAEA-Review-2024-049.pdf:13.94MB

In the aftermath of the Fukushima Daiichi Nuclear Power Station accident, safety measures against hydrogen in severe accident has been recognized as a serious technical problem in Japan. As one of efforts to form a common knowledge base between nuclear engineers and experts on combustion and explosion, we issued the "Handbook of Advanced Nuclear Hydrogen Safety (1st edition)" in 2017. For improvement of the rational advancement of hydrogen safety measures and further reliability of hydrogen safety evaluation, a CFD analysis is highly expected to produce more precisely and quantitative results. We have been developing an integrated CFD analysis code system which can analyze hydrogen diffusion, explosion-combustion and structural integrity at the severe accident especially for pressurized water reactors (PWRs). We organized the role of LP and the CFD analyses and their utilization examples of hydrogen safety validation. Based on these results, we made the "Handbook of Advanced Nuclear Hydrogen Safety (2nd volume)". The analysis results of real scale PWR described in 2nd volume are confirmed by cross-analysis models and existing data obtained through representative small, medium and large-scale tests.

Development of a new crust model for analyzing VULCANO VBS-U3 mcci experiment with MPS method

Yamada, Takeshi*; Li, X.; Yamashita, Takuya; Yamaji, Akifumi*

Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 10 Pages, 2024/11

https://doi.org/10.1115/ICONE31-134499

In this study, a new crust model is being developed to analyze MCCI, which involves continuous concrete ablation with presence of the crust layer between the corium and the concrete walls, which may gradually move with the slow concrete wall ablation process over long time. The new crust model must enable accumulation of physical displacement of the crust particle over long time (i.e., enable physical creeping) while preventing accumulation of numerical displacement of the crust particles over long time (i.e., preventing numerical creeping), Hence, in the new crust model, the PS has been effectively disabled for the crust particles. Qualitative validity of such numerical modeling was confirmed through some trial analyses of VULCANO-VBS test using a set of tentative calculation conditions and parameters, which should be carefully revised for future quantitative discussions including validation of the analysis results with experimental results.

France-Japan collaboration on severe accident studies in sodium-cooled fast reactors, 2; Methodologies and calculations of severe accident phases

Sogabe, Joji; Ishida, Shinya; Tagami, Hirotaka; Okano, Yasushi; Kamiyama, Kenji; Onoda, Yuichi; Matsuba, Kenichi; Yamano, Hidemasa; Kubo, Shigenobu; Kubota, Ryuzaburo*; et al.

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

In the frame of France-Japan collaboration, the calculational methodologies were defined and assessed, and the phenomenology and the severe accident consequences were investigated in a pool-type sodium-cooled fast reactor.

Experimental determination of deposition velocity of CsOH aerosols on CaCO at temperature range 170 - 290C

Luu, V. N.; Nakajima, Kunihisa

Nuclear Engineering and Design, 426, p.113402_1 - 113402_7, 2024/09

https://doi.org/10.1016/j.nucengdes.2024.113402

Fukushima Daiichi Nuclear Power Plant accident analysis considering the thermal stratification and containment leakage

Nakamura, Yuki*; Kojima, Yoshihiro*; Yamashita, Takuya; Shimomura, Kenta; Mizokami, Shinya

Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NUTHOS-14) (Internet), 12 Pages, 2024/08

Numerical study of initiating phase of core disruptive accident in small sodium-cooled fast reactors with negative void reactivity

Ishida, Shinya; Fukano, Yoshitaka; Tobita, Yoshiharu; Okano, Yasushi

Journal of Nuclear Science and Technology, 61(5), p.582 - 594, 2024/05

https://doi.org/10.1080/00223131.2023.2243949