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Sugita, Yutaka; Taniguchi, Naoki; Makino, Hitoshi; Kanamaru, Shinichiro*; Okumura, Taisei*
Nihon Genshiryoku Gakkai Wabun Rombunshi, 19(3), p.121 - 135, 2020/09
A series of structural analysis of disposal containers for direct disposal of spent fuel was carried out to provide preliminary estimates of the required pressure resistance thickness of the disposal container. Disposal containers were designed to contain either 2, 3 or 4 spent fuel assemblies in linear, triangular or square arrangements, respectively. The required pressure resistance thickness was evaluated using separation distance of the housing space for each spent fuel assembly as a key model parameter to obtain the required thickness of the body and then the lid of the disposal container. This work also provides additional analytical technical knowledge, such as the validity of the setting of the stress evaluation line and the effect of the model length on the analysis. These can then be referred to and used again in the future as a basis for conducting similar evaluations under different conditions or proceeding with more detailed evaluations.
Li, X.; Sato, Ikken; Yamaji, Akifumi*; Duan, G.*
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 8 Pages, 2018/07
Molten corium-concrete interaction (MCCI) is an important ex-vessel phenomenon that could happen during the late phase of a hypothetical severe accident in a light water reactor. In the present study, a three-dimensional (3-D) numerical study has been performed to simulate COMET-L3 test carried out by KIT with a stratified molten pool configuration of simulant materials with improved MPS method. The heat transfer between corium/crust/concrete was modeled with heat conduction between particles. Moreover, the potential influence of the siliceous aggregates was also investigated by setting up two different case studies since there was previous study indicating that siliceous aggregates in siliceous concrete might contribute to different axial and radial concrete ablation rates. The simulation results have indicated that metal melt as corium in MCCI can have completely different characteristics regarding concrete ablation pattern from that of oxidic corium, which needs to be taken into consideration when assessing the containment melt-through time in severe accident management.
Kitamura, Yasunori*; Fukushima, Masahiro
Nuclear Science and Engineering, 186(2), p.168 - 179, 2017/05
Times Cited Count：1 Percentile：10.58(Nuclear Science & Technology)An inconsistency between the reactivity worth of short-size samples measured by the critical-water-level (CWL) method and that conventionally analysed for validating the nuclear data and the nuclear calculation methods has been known. The present study investigated this inconsistency in terms of a simple theoretical framework and proposed a simple and practical technique for correcting the measured sample reactivity worth without making supplementary experiments. A series of Monte Carlo calculations that simulated typical sample reactivity worth measurement by the CWL method showed that this inconsistency is effectively reduced by the present correction technique.
Uchibori, Akihiro; Ohshima, Hiroyuki
Mechanical Engineering Journal (Internet), 3(3), p.15-00620_1 - 15-00620_9, 2016/06
For assessment of the wastage environment under tube failure accident, a mechanistic computer code called SERAPHIM calculating compressible multicomponent multiphase flow with sodium-water chemical reaction has been developed. In this study, applicability of the SERAPHIM code was investigated through the analysis of the experiment on water vapor discharging in liquid sodium under actual condition of the steam generator. The numerical result showed that the underexpanded jet appeared and impinged on the target tube located above the discharging tube. The calculated temperature distribution agreed with the measurement result well. The liquid droplet entrainment and its transport were considered in this analysis. The region with higher impingement velocity of the liquid droplet was close to the wastage region confirmed in the experiment. It was demonstrated that the SERAPHIM code could predict the temperature distribution and the environment of LDIE under the actual condition.
Guo, L.*; Morita, Koji*; Tobita, Yoshiharu
Journal of Nuclear Science and Technology, 53(2), p.271 - 280, 2016/02
Times Cited Count：7 Percentile：54.81(Nuclear Science & Technology)Uchibori, Akihiro; Ohshima, Hiroyuki
Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.533 - 544, 2015/08
For assessment of the wastage environment under tube failure accident, a mechanistic computer code called SERAPHIM calculating compressible multicomponent multiphase flow with sodium-water chemical reaction has been developed. In this study, applicability of the SERAPHIM code including the numerical model for liquid droplet entrainment and transport was investigated through the analysis of the basic experiment and the experiment under actual condition of the steam generator. In the analysis of the basic experiment, the calculated pressure variation during liquid droplet entrainment was consistent with the experimental result. In the analysis of the actual condition, the calculated temperature distribution agreed with the measurement result well. The region with higher impingement velocity of the liquid droplet was close to the wastage region confirmed in the experiment. It was demonstrated that the SERAPHIM code could predict the wastage environment under the actual condition.
Uchibori, Akihiro; Ohshima, Hiroyuki
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05
For assessment of the wastage environment under tube failure accident, a mechanistic computer code called SERAPHIM calculating compressible multicomponent multiphase flow with sodium-water chemical reaction has been developed. In this study, applicability of the SERAPHIM code was investigated through the analysis of the experiment on water vapor discharging in liquid sodium under actual condition of the steam generator. The numerical result showed that the underexpanded jet appeared and impinged on the target tube located above the discharging tube. The calculated temperature distribution agreed with the measurement result well. The liquid droplet entrainment and its transport were considered in this analysis. The region with higher impingement velocity of the liquid droplet was close to the wastage region confirmed in the experiment. It was demonstrated that the SERAPHIM code could predict the temperature distribution and the environment of LDIE under the actual condition.
Yoshida, Hiroyuki; Tamai, Hidesada; Onuki, Akira; Takase, Kazuyuki; Akimoto, Hajime
Nuclear Engineering and Technology, 38(2), p.119 - 128, 2006/04
The reduced-moderation water reactor core adopts a hexagonal tight-lattice arrangement. In the core, there is no sufficient information about the effects of the gap spacing and grid spacer configuration on the flow characteristics. Thus, we start to develop a predictable technology for thermal-hydraulic performance of the core using an advanced numerical simulation technology. As a part of this technology development, we are developing a two-phase flow simulation code TPFIT with an advanced interface tracking method. The vector and parallelization of the code was conducted to fit the large-scale simulations. The numerical results applied to large-scale water-vapor two-phase flow in tight lattice rod bundles are shown and compared with experimental results. In the results, a tendency of the predicted void fraction distribution in horizontal plane agreed with the measured values including the bridge formation of the liquid at the position of adjacent fuel rods where an interval is the narrowest.
Onuki, Akira; Takase, Kazuyuki; Kureta, Masatoshi*; Yoshida, Hiroyuki; Tamai, Hidesada; Liu, W.; Akimoto, Hajime
Proceedings of Japan-US Seminar on Two-Phase Flow Dynamics, p.317 - 325, 2004/12
We start R&D project to develop the predictable technology for thermal-hydraulic performance of Reduced-Moderation Water Reactor (RMWR) in collaboration with Power Company/reactor vendor/university since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured BWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron energy. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R&D items for the RMWR because of the tight lattice configuration. In this paper, we will show the R&D plan and describe the current status on experimental and analytical studies. We will confirm the thermal-hydraulic performance in the tight-lattice bundles by this project and develop a predictable technology for the RMWR in future.
Utsumi, Takayuki*; Kimura, Hideo
JSME International Journal, Series B, 47(4), p.761 - 767, 2004/11
In this paper, we show that a new numerical method, the Constrained Interpolation Profile - Basis Set (CIP-BS) method, can solve partial differential equations (PDEs) with high accuracy and can be a universal solver by presenting examples for the solutions of typical parabolic, hyperbolic, and elliptic equations. Here, we present the numerical errors caused by this method, and illustrate that the solutions by the CIP-BS
method, in which fifth order polynomials are used to constrain the values and first and second order spatial derivatives, are highly refined compared to those by the CIP-BS
method, in which third order polynomials are used to constrain the values and first order spatial derivatives. The fact that this method can unambiguously solve PDEs with an one-to-one correspondence to analytical requirements is also shown for PDEs including singular functions like the Dirac delta function with Dirichet or Neumann boundary conditions. This method is straightforwardly applicable to PDEs describing complex physical and engineering problems.
Yoshida, Hiroyuki; Nagayoshi, Takuji*; Ose, Yasuo*; Takase, Kazuyuki; Akimoto, Hajime
Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(3), p.233 - 241, 2004/09
When there are no experimental data such as the reduced-moderation water reactor (RMWR), therefore, it is very difficult to obtain highly precise predictions. The RMWR core adopts a hexagonal tight lattice arrangement with about 1 mm gap between adjacent fuel rods. In the core, there is no sufficient information about the effects of the gap spacing and grid spacer configuration on the flow characteristics. Thus, we start to develop a predictable technology for thermal-hydraulic performance of RMWR core using advanced numerical simulation technology. As part of this technology development, we are developing advanced interface tracking method to improve conservation of volume of fluid. In this paper, we describe a newly developed interface tracking method and examples of the numerical results. In the present results, the error of volume conservation in the bubbly flow is within 0.6%.
Soukhovitskij, E. Sh.*; Morogovskij, G. B.*; Chiba, Satoshi; Iwamoto, Osamu; Fukahori, Tokio
JAERI-Data/Code 2004-002, 32 Pages, 2004/03
JAERI-Data-Code-2004-002.pdf:1.43MB
This report gives a detailed description of the theory and computational algorithms of modernized coupled-channels optical model code OPTMAN based on the soft-rotator model for the collective nuclear structure and excitations. This work was performed under the Project Agreement B-521 with the International Science and Technology Center (Moscow), financing party of which is Japan. As a result of this work, the computational method of OPTMAN was totally updated, and an user-friendly interface was attached.
Koshizuka, Seiichi*; Ikeda, Hirokazu*; Liu, J.*; Oka, Yoshiaki*
JAERI-Tech 2002-013, 60 Pages, 2002/03
JAERI-Tech-2002-013.pdf:3.16MB
no abstracts in English
Yabe, Takashi*; Utsumi, Takayuki*
Computational Fluid Dynamics Journal, 9(3), p.185 - 193, 2000/10
no abstracts in English
Utsumi, Takayuki*; Sasaki, Akira; Fujii, Sadao*
Keisan Kogaku Koenkai Rombunshu, 5, 4 Pages, 2000/05
no abstracts in English
Utsumi, Takayuki*; Sasaki, Akira
Dai-14-Kai Suchi Ryutai Rikigaku Shimpojiumu Koen Yoshishu, 4 Pages, 2000/00
no abstracts in English
*; Kunugi, Tomoaki
Nihon Kikai Gakkai Rombunshu, B, 64(617), p.65 - 70, 1998/01
no abstracts in English
JAERI-Research 97-080, 43 Pages, 1997/11
JAERI-Research-97-080.pdf:1.24MB
no abstracts in English
JAERI-Research 97-079, 57 Pages, 1997/11
JAERI-Research-97-079.pdf:1.63MB
no abstracts in English
Yamazawa, Hiromi
Journal of Nuclear Science and Technology, 33(1), p.69 - 77, 1996/01
Times Cited Count：5 Percentile：45.07(Nuclear Science & Technology)no abstracts in English
