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Yamashita, Susumu; Tada, Kenichi; Yoshida, Hiroyuki; Suyama, Kenya
Nippon Genshiryoku Gakkai Wabun Rombunshi, 17(3/4), p.99 - 105, 2018/12
In order to reveal melt relocation behaviors of core internals phenomenologically and to reduce the uncertainties of the melt relocation analysis in existing SA analysis codes, in JAEA, the numerical simulation code for melt relocation and accumulation behaviors based on computational fluid dynamics named JUPITER has been developed. In this paper, to consider the estimation method for fuel debris composition and its re-criticality, we performed the melt accumulating and spreading simulation to the pedestal region by JUPITER and also performed re-criticality analysis by Monte Carlo Codes for Neutron Transport Calculations based on Continuous Energy and Multi-group Methods (MVP) using detailed fuel debris composition data obtained by JUPITER. From the coupled analysis on fuel debris distribution by JUPITER and MVP, we had prospects for a detailed possibility of re-criticality of fuel debris with detailed fuel debris distribution.
Uesawa, Shinichiro; Yamashita, Susumu; Shibata, Mitsuhiko; Yoshida, Hiroyuki
Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 6 Pages, 2018/11
Yamashita, Susumu; Yoshida, Hiroyuki
Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 5 Pages, 2018/11
To investigate detailed flow behaviors around the beam window of accelerator driven system (ADS), large scale simulation for unsteady thermal hydraulics around the beam window was performed using JUPITER. The input data, such as the beam window and nozzle, is designed as accurate as possible using the computer aided design software. As a result, the flow pattern around the beam window is quite different from previous results in which the steady flow is assumed. The flow pattern of the Lead-Bismuth Eutectic around the beam window and the exit of the nozzle were very complicated. According to complicated flow around those structures, the temperature distribution on the beam window is also complicated. Thus, it is confirmed that the complicated flow around structures will affect to the temperature distribution in the structures and the effect of flow field on the temperature must be evaluated.
Idomura, Yasuhiro; Ina, Takuya*; Yamashita, Susumu; Onodera, Naoyuki; Yamada, Susumu; Imamura, Toshiyuki*
Proceedings of 9th Workshop on Latest Advances in Scalable Algorithms for Large-Scale Systems (ScalA 2018) (Internet), p.17 - 24, 2018/11
A communication avoiding (CA) multigrid preconditioned conjugate gradient method (CAMGCG) is applied to the pressure Poisson equation in a multiphase CFD code JUPITER, and its computational performance and convergence property are compared against CA Krylov methods. In the JUPITER code, the CAMGCG solver has robust convergence properties regardless of the problem size, and shows both communication reduction and convergence improvement, leading to higher performance gain than CA Krylov solvers, which achieve only the former. The CAMGCG solver is applied to extreme scale multiphase CFD simulations with 
billion DOFs, and it is shown that compared with a preconditioned CG solver, the number of iterations is reduced to 
, and 
speedup is achieved with keeping excellent strong scaling up to 8,000 nodes on the Oakforest-PACS.
Uesawa, Shinichiro; Yamashita, Susumu; Shibata, Mitsuhiko; Yoshida, Hiroyuki
Mechanical Engineering Journal (Internet), 5(4), p.18-00115_1 - 18-00115_13, 2018/08
Yamashita, Susumu; Yoshida, Hiroyuki
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 6 Pages, 2018/07
Yamashita, Susumu; Ina, Takuya*; Idomura, Yasuhiro; Yoshida, Hiroyuki
Dai-31-Kai Suchi Ryutai Rikigaku Shimpojiumu Koen Rombunshu (DVD-ROM), 7 Pages, 2017/12
no abstracts in English
Yamashita, Susumu; Ina, Takuya; Idomura, Yasuhiro; Yoshida, Hiroyuki
Nuclear Engineering and Design, 322, p.301 - 312, 2017/10
Times Cited Count：4 Percentile：18.09(Nuclear Science & Technology)In recent years, significant attention has been paid to the precise determination of relocation of molten materials in reactor pressure vessels of boiling water reactors (BWRs) during severe accidents. To address this problem, we have developed a computational fluid dynamics code JUPITER, based on thermal-hydraulic equations and multi-phase simulation models. Although the Poisson solver has previously been a performance bottleneck in the JUPITER code, this is resolved by a new hybrid parallel Poisson solver, whose strong scaling is extended up to 
200k cores on the K-computer. As a result of the improved computational capability, the problem size and physical models are dramatically expanded. A series of verification and validation studies are enabled, which are in agreement with previous numerical simulations and experiments. These physical and computational capabilities of JUPITER enable us to investigate molten material behaviors in reactor relevant situations.
Yamashita, Susumu; Uesawa, Shinichiro; Yoshida, Hiroyuki
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 8 Pages, 2017/07
Uesawa, Shinichiro; Yamashita, Susumu; Shibata, Mitsuhiko; Yoshida, Hiroyuki
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 7 Pages, 2017/07
Yamashita, Susumu; Tokushima, Kazuyuki*; Kurata, Masaki; Yoshida, Hiroyuki
Mechanical Engineering Journal (Internet), 4(3), p.16-00567_1 - 16-00567_13, 2017/06
In order to precisely investigate molten core relocation behavior in severe accidents, we have been developing the detailed and phenomenological numerical simulation code named JUPITER for predicting the molten core behavior with melting and solidification based on computational fluid dynamics (CFD) including the three-dimensional multiphase thermal-hydraulic simulation models. In order to treat complicated core structures, e.g., boron carbide (absorber), stainless steel (control rod, fuel support structure, etc.), Zircaloy (channel box and fuel cladding) and to deal with complicated melt relocation behaviors, high accuracy, efficient, stable and robust numerical schemes are implemented. In this paper, in order to evaluate the validity and applicability of the JUPITER for actual core structures, we perform the preliminary melt relocation analysis in the control rod and fuel support piece and also verify the validity of the JUPITER regarding the melt relocation and solidification processes by the fundamental numerical problem and the experimental analysis. As a result, the preliminary analysis showed that multicomponent melt flow and its melt and solidification were stably worked in the melt relocation simulation. In the validation analysis, the numerical results were in the reasonably agreement with experimental results. Therefore, it was confirmed that the JUPITER has a potential to calculate the core melt relocation behavior in RPVs.
Yamashita, Susumu; Uesawa, Shinichiro; Yoshida, Hiroyuki
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04
In order to precisely and unified investigate molten core relocation behavior in the Fukushima Daiichi Nuclear Power Station, we have developed the detailed and phenomenological numerical simulation code named JUPITER for predicting the molten core behavior including solidification and relocation based on the three-dimensional multiphase thermal-hydraulic simulation models. At the moment, the fundamental framework for melt relocation behavior and natural convection around accumulated debris in the JUPITER has been developed, that is, core melt and its relocation, corium spreading behavior in a pedestal region, simulation of air cooling evaluation method. In this paper, preliminary analyses, e.g., numerical simulation for core melt relocation behavior, corium spreading behavior and air cooling analysis around debris are shown.
Yamashita, Susumu; Uesawa, Shinichiro; Yoshida, Hiroyuki
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11
In fuel debris retrieval in decommissioning of the Fukushima Daiichi NPS, dry method is under consideration. Investigation of the cooling performance of fuel debris in the dry method will be very important problem to realize the method. However, there are uncertainties in the shape and surface temperature of fuel debris. In order to evaluate the cooling performance, the investigation of the cooling performance by free convection is required. We have been developing the numerical simulation method, which can evaluate the cooling performance of the fuel debris by free convection, using the JUPITER code in JAEA. In this paper, we show the evaluation result of the thermal conductivity by the free convection from fuel debris to the atmosphere in the simplified system.
Uesawa, Shinichiro; Shibata, Mitsuhiko; Yamashita, Susumu; Yoshida, Hiroyuki
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11
Yoshida, Hiroyuki; Uesawa, Shinichiro; Yamashita, Susumu; Nagase, Fumihisa
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11
Yamashita, Susumu; Tokushima, Kazuyuki; Kurata, Masaki; Takase, Kazuyuki; Yoshida, Hiroyuki
Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 5 Pages, 2016/06
In order to precisely investigate molten core relocation behavior in the Fukushima Daiichi Nuclear Power Station, we have developed the detailed and phenomenological numerical simulation code named JUPITER for predicting the molten core behavior including solidification and relocation based on the three-dimensional multiphase thermal-hydraulic simulation models. At the moment, multicomponent analysis method which can be treated any number of component as a fluid or solid body, Zr-water reaction model and simple radiation heat transfer model were implemented and showed that multicomponent melt flow and its solidification were confirmed in the simplified core structure system. However, the validation of the JUPITER using high temperature molten material has not been performed yet. In this paper, in order to evaluate the validity of the JUPITER, especially, for high temperature melt relocation experiment, we compared between numerical and experimental results for that system. As a result, qualitatively reasonable result was obtained. And also we performed melt relocation simulation on actual core structures designed by three dimensional CAD (Computer-Aided Design) and then we estimated phenomena which might be actually occurred in SAs.
Yamashita, Susumu; Tokushima, Kazuyuki; Kurata, Masaki; Takase, Kazuyuki; Yoshida, Hiroyuki
Nippon Kikai Gakkai Dai-28-Kai Keisan Rikigaku Koenkai Rombunshu (CD-ROM), 3 Pages, 2015/10
no abstracts in English
Yamashita, Susumu; Takase, Kazuyuki; Yoshida, Hiroyuki
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05
In accidents of the Fukushima Daiichi Nuclear Power Plants, by stop of the emergency core cooling system, fuel rods were overheated due to the radioactive decay heat and the oxidization of fuel cladding. Although it is inferred that the core degradation occurred, condition inside the core still has not been revealed. Especially, in order to precisely understand the accumulation condition of debris in lower plenum, detailed and phenomenological relocation process of molten fuel is quite important. In this problem, since an experiment is extremely difficult, numerical simulation will be useful tool for investigating conditions in reactor core. However, existing codes can not be phenomenologically treated relocations process. Therefore a phenomenologically-based numerical simulation method for predicting the melting core behavior including solidification and relocation based on the computational fluid dynamics has been developed in JAEA. Last paper, ICONE 23, in order to distinguish a fuel component which possesses a heat source such as a decay heat from structures which does not possess a heat source such as a core plate, control rod guide tubes and so on, we developed three phases and three component multiphase flow simulation code and performed preliminary analysis of molten core relocation behavior using simplified core structures. As a result, we obtained reasonable results. However, we have not carried out validations for the numerical code yet. In this paper, we show that the results of the numerical test for evaluating the validity of the numerical code and also show that the development of the radiation heat transfer model and its preliminary analysis. In addition, we will report the preliminary analysis of the relocation behavior of molten materials in the simplified core support structure.
Yamashita, Susumu; Takase, Kazuyuki; Yoshida, Hiroyuki
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 9 Pages, 2014/07
In accidents of the Fukushima Daiichi Nuclear Power Plants, by stop of the emergency core cooling system, fuel rods were overheated due to the radioactive decay heat and the oxidization of fuel cladding. Although it is inferred that the core degradation occurred because fuels, control rods, and other components in a reactor vessel was melted and relocated, condition inside the core still has not been revealed. Especially, in order to precisely understand the accumulation condition of debris in lower plenum, detailed and phenomenological relocation process of molten fuel is quite important. In this problem, since an experiment is extremely difficult, numerical simulation will be useful tool for investigating conditions in reactor core. However, existing codes can not be phenomenologically treated relocations process. In order to correctly estimate progress of the relocation phenomena in the reactor core, a numerical simulation code that can phenomenologically evaluate the melting phenomena is required. Therefore a phenomenologically-based numerical simulation method for predicting the melting core behavior including solidification and relocation based on the computational fluid dynamics has been developed in JAEA. Last paper, ICONE 21, we carried out the calculation of relocation behavior using three phase (solid/liquid/gas) and two components (metal and gas) fluid flow simulation model, however, there is only one component for metal. Therefore, the model cannot distinguish fuel material with decay heat from core internal materials. In this paper, we show the brief overview about the extended code, which is added one more component to the previous code to distinguish a fuel material with constant heat source simulating decay heat in the energy equation from core internals, and also show that the numerical results of relocation behavior for molten fuel and core internals in a reactor core.
Yamashita, Susumu; Takase, Kazuyuki
Dai-19-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.153 - 156, 2014/06
no abstracts in English
