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Motegi, Kosuke; Shiotsu, Hiroyuki; Matsumoto, Toshinori; Hibiki, Takashi*; Shibamoto, Yasuteru
International Journal of Heat and Mass Transfer, 258, p.128275_1 - 128275_15, 2026/05
Times Cited Count：2 Percentile：0.00(Thermodynamics)Wang, Z.; Matsumoto, Toshinori; Shibamoto, Yasuteru; Duan, G.*
Journal of Computational Physics, 537, p.114072_1 - 114072_29, 2025/09
Times Cited Count：2 Percentile：70.80(Computer Science, Interdisciplinary Applications)Motegi, Kosuke; Matsumoto, Toshinori; Shiotsu, Hiroyuki
Proceedings of 21st International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-21) (Internet), 10 Pages, 2025/08
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.
Matsumoto, Toshinori; Hibiki, Takashi*; Maruyama, Yu
International Journal of Energy Research, 2024(1), p.9748588_1 - 9748588_18, 2024/08
Times Cited Count：1 Percentile：20.46(Energy & Fuels)To evaluate the effectiveness of the wet cavity strategy, the authors developed a stochastic evaluation method that considers the uncertainties of the molten material conditions ejected from reactor pressure vessels. The first step was uncertainty analysis using the MELCOR code to obtain the melt condition. Five uncertain parameters related to the core degradation process were chosen. The input parameter sets were generated using Latin hypercube sampling. The second step was analyzing of the melt-behavior using the JASMINE code. The probabilistic distribution of parameters for the JASMINE analyses was determined from the MELCOR analysis results. The initial water depth was set to 0.5, 1.0, and 2.0 m. The debris height was compared with the criterion to judge its coolability. Consequently, the success probability of debris cooling was obtained through a sequence of calculations. The feasibility and technical difficulties in the MELCOR-JASMINE combined analysis were also discussed.
Sahboun, N. F.; Matsumoto, Toshinori; Iwasawa, Yuzuru; Wang, Z.; Sugiyama, Tomoyuki
Annals of Nuclear Energy, 195, p.110145_1 - 110145_12, 2024/01
Times Cited Count：3 Percentile：40.29(Nuclear Science & Technology)Wang, Z.; Matsumoto, Toshinori; Duan, G.*; Matsunaga, Takuya*
Computer Methods in Applied Mechanics and Engineering, 414, p.116168_1 - 116168_49, 2023/09
Times Cited Count：32 Percentile：95.29(Engineering, Multidisciplinary)Matsumoto, Toshinori; Kawabe, Ryuhei*; Iwasawa, Yuzuru; Sugiyama, Tomoyuki; Maruyama, Yu
Annals of Nuclear Energy, 178, p.109348_1 - 109348_13, 2022/12
Times Cited Count：3 Percentile：30.55(Nuclear Science & Technology)The Japan Atomic Energy Agency extended the applicability of their fuel-coolant interaction analysis code JASMINE to simulate the relevant phenomena of molten core in a severe accident. In order to evaluate the total coolability, it is necessary to know the mass fraction of particle, agglomerated and cake debris and the final geometry at the cavity bottom. An agglomeration model that considers the fusion of hot particles on the cavity floor was implemented in the JASMINE code. Another improvement is introduction of the melt spreading model based on the shallow water equation with consideration of crust formation at the melt surface. For optimization of adjusting parameters, we referred data from the agglomeration experiment DEFOR-A and the under-water spreading experiment PULiMS conducted by KTH in Sweden. The JASMINE analyses reproduced the most of the experimental results well with the common parameter set, suggesting that the primary phenomena are appropriately modelled.
Sahboun, N. F.; Matsumoto, Toshinori; Iwasawa, Yuzuru; Sugiyama, Tomoyuki
Proceedings of Asian Symposium on Risk Assessment and Management 2021 (ASRAM 2021) (Internet), 15 Pages, 2021/10
Matsumoto, Toshinori; Iwasawa, Yuzuru; Sugiyama, Tomoyuki
Proceedings of Reactor core and Containment Cooling Systems, Long-term management and reliability (RCCS 2021) (Internet), 8 Pages, 2021/10
A methodological framework is being developed in JAEA for evaluating debris coolability at ex-vessel during the severe accident (SA) of BWR under the wet cavity strategy. The probability of ex-vessel debris coolability under the wet cavity strategy is analyzed to demonstrate the evaluation approach. Probabilistic distribution of the melt conditions ejected from the RPV was obtained as the result of the iterative analyses with MELCOR code. Five uncertainty parameters relating with the core degradation and transfer process were chosen. Parameter sets were generated by Latin hypercube sampling (LHS). JASMINE code plays the physical model to predict the mass fraction of agglomerated debris and melt pool spreading on the floor. Fifty-nine input parameter set for JASMINE code were generated by LHS again using the probabilistic distribution of melt condition determined from the results of MELCOR analyses. The depth of the water pool was set as 0.5, 1.0 and 2.0 m. The accumulated debris height was compared with the criterion to judge the debris coolability. As the result, the success probability of debris cooling was obtained through the sequence of calculations.
Matsumoto, Toshinori; Iwasawa, Yuzuru; Ajima, Kohei*; Sugiyama, Tomoyuki
Proceedings of Asian Symposium on Risk Assessment and Management 2020 (ASRAM 2020) (Internet), 10 Pages, 2020/11
The probability of ex-vessel debris coolability under the wet cavity strategy is analyzed. The first step is the uncertainty analyses by severe accident analysis code MELCOR to obtain the melt condition. Five uncertain parameters which are relating with the core degradation and transfer process were chosen. Input parameter sets were generated by LHS. The analyses were conducted and the conditions of the melt were obtained. The second step is the analyses for the behavior of melt under the water by JASMINE code. The probabilistic distribution of parameters are determined from the results of MELCOR analyses. Fifty-nine parameter sets were generated by LHS. The depth of water pool is set to be 0.5, 1.0 and 2.0 m. Debris height were compared with the criterion to judge the debris coolability. As the result, the success probability of debris cooling was obtained through the sequence of calculations. The technical difficulties of this evaluation method are also discussed.
Hotta, Akitoshi*; Akiba, Miyuki*; Morita, Akinobu*; Konovalenko, A.*; Vilanueva, W.*; Bechta, S.*; Komlev, A.*; Thakre, S.*; Hoseyni, S. M.*; Sk
ld, P.*; et al.
Journal of Nuclear Science and Technology, 57(4), p.353 - 369, 2020/04
Times Cited Count：22 Percentile：72.57(Nuclear Science & Technology)Motegi, Kosuke; Trianti, N.; Matsumoto, Toshinori; Sugiyama, Tomoyuki; Maruyama, Yu
Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.4324 - 4335, 2019/08
Matsumoto, Toshinori; Sato, Masatoshi; Sugiyama, Tomoyuki; Maruyama, Yu
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 6 Pages, 2017/07
Sato, Masatoshi; Matsumoto, Toshinori; Sugiyama, Tomoyuki; Maruyama, Yu
Proceedings of 8th European Review Meeting on Severe Accident Research (ERMSAR 2017) (Internet), 12 Pages, 2017/05
Sato, Masatoshi; Matsumoto, Toshinori; Sugiyama, Tomoyuki; Maruyama, Yu
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 10 Pages, 2016/11
Matsumoto, Toshinori; Kawabe, Ryuhei; Sugiyama, Tomoyuki; Maruyama, Yu
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 9 Pages, 2016/11
During severe accident at nuclear power stations, molten core material jet could be discharged from the reactor pressure vessel into the water pool formed at the pedestal or cavity in the containment vessel. To improve the JASMINE code, The method for determining particle diameters which follow the Rosin-Rammler distribution was implemented. The jet breakup experiments, DEFOR-A conducted by KTH (Royal Institute of Technology, Sweden) were analyzed with the code. The influence of the experimental conditions, such as water subcooling, melt jet diameter and superheat were discussed. A crust layer formation model was also implemented in the code. The analyses using the model were carried out for the melt spreading experiments, PULiMS conducted by KTH. The spreading area was overestimated. Further improvement of the melt spreading model were discussed to close the gaps by introducing additional models such as heat conduction in the substrate materials, void formed inside the melt and so on.
Matsumoto, Toshinori; Ishikawa, Jun; Maruyama, Yu
Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.4033 - 4043, 2015/08
Matsumoto, Toshinori*; Takata, Takashi*; Yamaguchi, Akira*; Kurihara, Akikazu; Ohshima, Hiroyuki
Journal of Nuclear Science and Technology, 48(3), p.315 - 321, 2011/03
Times Cited Count：1 Percentile：9.92(Nuclear Science & Technology)In the steam generator of a sodium-cooled fast reactor, when the tube fails, water leaks into the sodium stream and sodium-water reaction is initiated. In the present study, a numerical analysis has been carried out to determine the heat transfer coefficient from temperature data obtained in a sodium-water reaction experiment. By updating the heat transfer coefficient, an inverse problem of heat transfer has been solved in the analysis based on the result of the SWAT-1R experiment. It is found that the heat transfer coefficient fluctuates largely during the reaction. The heat transfer coefficient is affected by the flow characteristics. Hence, we characterize the flow pattern near the heat transfer tube at typical periods in the phenomenon progression.
Matsumoto, Toshinori*; Takata, Takashi*; Yamaguchi, Akira*; Kurihara, Akikazu; Ohshima, Hiroyuki
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 10 Pages, 2009/09
When a heat transfer tube fails in steam generator (SG) of sodium-cooled fast reactor (SFR), sodium-water reaction (SWR) would take place. It is significant for estimation of the heat transfer from the fluid to the tube wall during SWR region to evaluate the possibility of the secondary tube failure in case of overheating rupture. In the present study, thermal hydraulics simulation of the fluid around the tube is conducted. The heat transfer coefficient is computed, the correlation diagram between the heat transfer coefficient and the void fraction has been obtained. The void fraction around the heat transfer tube in the SWR has been evaluated.
