Development of a formulation to predict molten core spreading in an LWR severe accident

Sahboun, N. F.; Matsumoto, Toshinori; Iwasawa, Yuzuru; Wang, Z.; Sugiyama, Tomoyuki

Annals of Nuclear Energy, 195, p.110145_1 - 110145_12, 2024/01

https://doi.org/10.1016/j.anucene.2023.110145

Improvement of JASMINE code for ex-vessel molten core coolability in BWR

Matsumoto, Toshinori; Kawabe, Ryuhei*; Iwasawa, Yuzuru; Sugiyama, Tomoyuki; Maruyama, Yu

Annals of Nuclear Energy, 178, p.109348_1 - 109348_13, 2022/12

https://doi.org/10.1016/j.anucene.2022.109348

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.

Development of evaluation framework for ex-vessel core coolability

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.

Experimental and analytical investigation of formation and cooling phenomena in high temperature debris bed

Hotta, Akitoshi*; Akiba, Miyuki*; Morita, Akinobu*; Konovalenko, A.*; Vilanueva, W.*; Bechta, S.*; Komlev, A.*; Thakre, S.*; Hoseyni, S. M.*; Skld, P.*; et al.

Journal of Nuclear Science and Technology, 57(4), p.353 - 369, 2020/04

https://doi.org/10.1080/00223131.2019.1691078

Fluid dynamic analysis on hydrogen deflagration in vertical flow channel with annular obstacles

Matsumoto, Toshinori; Sato, Masatoshi; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 6 Pages, 2017/07

Improvement of ex-vessel molten core behavior models for the JASMINE code

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.

Analysis for progression of accident at Fukushima Dai-ichi Nuclear Power Station with THALES2 code

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

Numerical study on correlation of heat transfer coefficient with void fraction at heat transfer tube surface in sodium water reaction

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.

Numerical investigation of heat transfer toward surrounding tube in sodium-water reaction phenomenon

Matsumoto, Toshinori*; Takata, Takashi*; Yamaguchi, Akira*; Uchibori, Akihiro; Ohshima, Hiroyuki

no journal, , 

no abstracts in English

Study on correlation of heat transfer coefficient with void fraction on heat transfer tube surface in sodium-water reaction

Matsumoto, Toshinori*; Takata, Takashi*; Yamaguchi, Akira*; Kurihara, Akikazu; Ohshima, Hiroyuki

no journal, , 

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.