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JAEA Reports

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.

Journal Articles

Analysis of ex-vessel debris coolability of boiling water reactors

Matsumoto, Toshinori; Hibiki, Takashi*; Maruyama, Yu

International Journal of Energy Research, 2024(1), p.9748588_1 - 9748588_18, 2024/08

 Times Cited Count:0 Percentile:0.00(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.

Journal Articles

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

 Times Cited Count:2 Percentile:14.04(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.

Journal Articles

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.

Journal Articles

The Analysis for Ex-Vessel debris coolability of BWR

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.

Journal Articles

Simulation of fuel-coolant interaction SERENA2 test based on JASMINE version 3

Hotta, Akitoshi*; Morita, Akinobu*; Kajimoto, Mitsuhiro*; Maruyama, Yu

Nihon Genshiryoku Gakkai Wabun Rombunshi, 16(3), p.139 - 152, 2017/09

Journal Articles

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.

Journal Articles

Analytical tool development for coarse break-up of a molten jet in a deep water pool

Moriyama, Kiyofumi; Nakamura, Hideo; Maruyama, Yu*

Nuclear Engineering and Design, 236(19-21), p.2010 - 2025, 2006/10

 Times Cited Count:25 Percentile:82.92(Nuclear Science & Technology)

A computer code JASMINE-pre was developed for the prediction of premixing conditions of fuel-coolant interactions and the debris bed formation behavior relevant to severe accidents of light water reactors. JASMINE-pre consists of three melt component models: melt jet, melt particles and melt pool, coupled with a two-phase flow model derived from the ACE-3D code developed at JAERI. Simulations of the FARO corium quenching experiments with a saturated water pool and with a subcooled water pool were performed with JASMINE-pre and ${tt pmjet}$. JASMINE-pre reproduced the pressurization and fragmentation behaviors observed in the experiments with a reasonable accuracy. The results by pmjet showed qualitatively the same trend with JASMINE-pre in the fragmentation behavior.

Journal Articles

Evaluation of ex-vessel steam explosion induced containment failure probability for Japanese BWR

Moriyama, Kiyofumi; Takagi, Seiji; Muramatsu, Ken; Nakamura, Hideo; Maruyama, Yu

Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), 9 Pages, 2005/05

The containment failure probability due to ex-vessel steam explosions were evaluated for a BWR Mk-II model plant. The evaluation was made for two scenarios: a steam explosion in the pedestal area, or in the suppression pool. A probabilistic approach, Latin Hypercube Sampling (LHS), was applied for the evaluation of steam explosion loads, in which a steam explosion simulation code JASMINE was used as a physics model. The fragility curves connecting the steam explosion loads and containment failure were developed based on simplified assumptions on the containment failure scenarios. The mean conditional probabilities of containment failure per occurrence of a steam explosion were $$6.4times 10^{-2}$$ for suppression pool and $$2.2times 10^{-3}$$ for pedestal area. Note that the results depend on the assumed range of input parameters and fragility curves that involve conservatism and simplification.

Journal Articles

Simulation of alumina and corium steam explosion experiments with JASMINE v.3

Moriyama, Kiyofumi; Nakamura, Hideo; Maruyama, Yu

Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 18 Pages, 2004/10

A steam explosion simulation code JASMINE is under development at JAERI for the assessment of steam explosion impacts on the integrity of containment vessel during severe accidents in light water reactors. Selected alumina and corium steam explosion experiments, KROTOS-44, 42, 37 and FARO-L33 were simulated with JASMINE code. The experimentally observed difference of the steam explosion intensity with the two materials, alumina and corium, was reproduced in the simulations without changing the model parameters related to the fine fragmentation process, but based on the difference in the premixing behavior predicted by the simulations. The simulation of corium experiments showed more fraction of the melt droplets frozen during premixing, as well as more void fraction, and those two points were likely to be the primary reasons of weak interactions in corium experiments.

JAEA Reports

JASMINE-pro: A Computer code for the analysis of propagation process in steam explosions; User's manual

Yang, Y.; Nilsuwankosit, S.; Moriyama, Kiyofumi; Maruyama, Yu; Nakamura, Hideo; Hashimoto, Kazuichiro

JAERI-Data/Code 2000-035, 86 Pages, 2000/12

JAERI-Data-Code-2000-035.pdf:4.01MB

no abstracts in English

Journal Articles

Study on premixing phase of steam explosion at JAERI

; Moriyama, Kiyofumi; Maruyama, Yu; H.Park*; Y.Yang*; Sugimoto, Jun

Nucl. Eng. Des., 189(1-3), p.205 - 221, 1999/00

 Times Cited Count:7 Percentile:49.07(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Study on premixing phase of steam explosion of JAERI

; Moriyama, Kiyofumi; Maruyama, Yu; H.Park*; Y.Yang*; Sugimoto, Jun

JAERI-Conf 97-011, p.447 - 466, 1998/01

no abstracts in English

Journal Articles

Study on steam explosion and molten core coolability in ALPHA program

Yamano, N.; Maruyama, Yu; Moriyama, Kiyofumi; Kudo, Tamotsu; H.S.Park*; Sugimoto, Jun

Proc. of 11th KAIF/KNS Annual Conf., 0, p.827 - 838, 1996/00

no abstracts in English

JAEA Reports

Development of steam explosion simulation code JASMINE

Moriyama, Kiyofumi; Yamano, N.; Maruyama, Yu; Kudo, Tamotsu; ; ; Sugimoto, Jun

JAERI-Data/Code 95-016, 50 Pages, 1995/11

JAERI-Data-Code-95-016.pdf:1.61MB

no abstracts in English

Journal Articles

Overview of vapor explosions in nuclear power safety

Sugimoto, Jun

Proc., Seminar on the Vapor Explosions in Nuclear Power Safety,Kanzanji 1995, 0, p.1 - 15, 1995/00

no abstracts in English

Oral presentation

Evaluation of molten core coolability in containment vessel using JASMINE code

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

no journal, , 

For assessment the molten core coolability during severe accident in LWRs, we improved the models in the JASMINE code regarding to agglomeration of melt particles and melt spreading in containment vessel based on the DEFOR-A and the PULiMS experiments conducted by KTH. The improved models generally show good agreement with these experimental results.

Oral presentation

Improvement of fuel-coolant interaction models for ex-vessel debris coolability evaluation

Matsumoto, Toshinori; Kawabe, Ryuhei; Sugiyama, Tomoyuki; Maruyama, Yu

no journal, , 

Oral presentation

Study on molten core coolability under precautionary water injection into containment vessel, 4; Improvement of jet breakup model in the JASMINE code

Matsumoto, Toshinori; Kawabe, Ryuhei; Sugiyama, Tomoyuki; Maruyama, Yu

no journal, , 

no abstracts in English

Oral presentation

Study on molten core coolability under precautionary water injection into containment vessel, 1; Development of breakup particle agglomeration model in JASMINE

Matsumoto, Toshinori; Kawabe, Ryuhei; Ajima, Kohei; Sugiyama, Tomoyuki; Maruyama, Yu

no journal, , 

no abstracts in English

22 (Records 1-20 displayed on this page)