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.

Development of a simplified boiling model applied for large-scale detailed two-phase flow simulations based on the VOF method

Ono, Ayako; Sakashita, Hiroto*; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki

Mechanical Engineering Journal (Internet), 11(4), p.24-00188_1 - 24-00188_12, 2024/07

https://doi.org/10.1299/mej.24-00188

Japan Atomic Energy Agency (JAEA) is developing the evaluation method for a two-phase flow in the reactor core using simulation codes based on the Volume Of Fluid (VOF) method. JAEA started developing a Simplified Boiling Model (SBM) for the large-scale two-phase flow in the fuel assemblies. In the SBM, the motion and growth equations of the bubble are solved to obtain their diameter and time length at the detachment, of which size scale is within/around the calculation grid size of the numerical simulation. JUPITER calculates the bubble behavior with a scale of more than several m. In this study, the convection boiling on a vertical heating surface is simulated using the developed SBM. The comparison between the simulation and experimental results showed good reproducibility of the heat flux and velocity dependency on the passage period of the bubble.

Survey on technical issues of fission products behavior for improvement of decommissioning work efficiency and source term predicting accuracy; Report on the activity of this research committee for 2 years

Katsumura, Kosuke*; Takagi, Junichi*; Hosomi, Kenji*; Miyahara, Naoya*; Koma, Yoshikazu; Imoto, Jumpei; Karasawa, Hidetoshi; Miwa, Shuhei; Shiotsu, Hiroyuki; Hidaka, Akihide*; et al.

Nihon Genshiryoku Gakkai-Shi ATOMO, 65(11), p.674 - 679, 2023/11

https://doi.org/10.3327/jaesjb.65.11_674

no abstracts in English

Double diffusive dissolution model of UO pellet in molten Zr cladding

Ito, Ayumi*; Yamashita, Susumu; Tasaki, Yudai; Kakiuchi, Kazuo; Kobayashi, Yoshinao*

Journal of Nuclear Science and Technology, 60(4), p.450 - 459, 2023/04

https://doi.org/10.1080/00223131.2022.2114956

Numerical simulation of annular dispersed flow in simplified subchannel of light water cooled fast reactor RBWR

Yoshida, Hiroyuki; Horiguchi, Naoki; Ono, Ayako; Furuichi, Hajime*; Katono, Kenichi*

Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 7 Pages, 2022/08

Numerical simulation of two-phase flow in fuel assemblies with a spacer grid using a mechanistically based method

Ono, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03

JAEA is developing the methodology to predict the critical heat flux based on a mechanism in order to reduce the cost for full mock-up test. The evaluation method based on a mechanism is expected to be able to predict in the wide range of parameter under the unexpected conditions including the severe accident. In this study, the JUPITER code developed by JAEA is examined to apply for the two-phase flow simulation of LWR fuel assembly with the spacer grid. The benchmark data of single-phase flow in the bundle with the spacers by KAERI were used to validate the simulation result by JUPITER. Moreover, the single-phase flow simulation was conducted by another simulation method, STAR-CCM+, as a supplemental analysis to consider the effect of the different simulation methods. Finally, the two-phase flow simulation for the bundle with the spacer was conducted by JUPITER. The effect of the spacer with a vane on the bubble behavior is discussed.

Development of accident tolerant FeCrAl-ODS fuel cladding for BWRs in Japan

Sakamoto, Kan*; Miura, Yusuke*; Ukai, Shigeharu; Ono, Naoko*; Kimura, Akihiko*; Yamaji, Akifumi*; Kusagaya, Kazuyuki*; Takano, Sho*; Kondo, Takao*; Ikegawa, Tomohiko*; et al.

Journal of Nuclear Materials, 557, p.153276_1 - 153276_11, 2021/12

https://doi.org/10.1016/j.jnucmat.2021.153276

A FeCrAl-oxide dispersion strengthened (ODS) alloy is a promising candidate alloy for the accident tolerant fuel (ATF) cladding of light water reactors (LWRs) and being developed in Japan recently. This paper will introduce the progress of development of accident tolerant FeCrAl-ODS fuel claddings for boiling water reactors (BWRs) in Japan. Both the experimental and the analytical studies have been performed to evaluate the influence of implementation of the FeCrAl-ODS fuel claddings to the current BWRs. The experimental studies have been conducted to obtain and accumulate key material properties of FeCrAl-ODS fuel claddings by using bar, sheet and tube-shaped materials to support the evaluations in the analytical studies. At the end of paper, the challenges and prospects found in the program are highlighted to enhance international collaborations to accelerate the development of FeCrAl-ODS fuel cladding.

Thermal-hydraulics to risk assessment; Roles of thermal-hydraulics simulation to risk assessment

Maruyama, Yu; Yoshida, Kazuo

Nihon Genshiryoku Gakkai-Shi ATOMO, 63(7), p.517 - 522, 2021/07

https://doi.org/10.3327/jaesjb.63.7_517

no abstracts in English

Numerical simulation of two-phase flow in 44 simulated bundle

Ono, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki

Mechanical Engineering Journal (Internet), 7(3), p.19-00583_1 - 19-00583_12, 2020/06

https://doi.org/10.1299/mej.19-00583

JAEA is implementing the 3D detailed nuclear-thermal-coupled analysis code to analyze the transition state of the core and to reduce the likelihood of the design. In the development plan, the computational fluid dynamics code based on the VOF method, JUPITER, is applied for TH part of the 3D detailed nuclear-thermal-coupled analysis code.

Irradiation growth behavior of improved Zr-based alloys for fuel cladding

Amaya, Masaki; Kakiuchi, Kazuo; Mihara, Takeshi

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.1048 - 1056, 2019/09