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Nakajima, Norihiro; Nishida, Akemi; Kawakami, Yoshiaki; Okada, Tatsuo*; Tsuruta, Osamu*; Sawa, Kazuhiro; Iigaki, Kazuhiko
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 9 Pages, 2014/07
Almost all industrial products are assembled from multiple parts. A nuclear facility is a large structure consisting of more than 10 million components. This paper discusses a method to analyze an assembly by gathering data on its component parts. Gathered data on component may identify ill conditioned meshes for connecting surfaces between components. These ill meshes are typified by nodal point disagreement in finite element discretization. A technique to resolve inconsistencies in data among the components is developed. By using this technique, structural analysis for an assembly can be carried out, and results can be obtained by the use of supercomputers, such as the K computer. Numerical results are discussed for components of the High Temperature Engineering Test Reactor.
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.
Ikusawa, Yoshihisa; Ozawa, Takayuki; Hirooka, Shun; Maeda, Koji; Kato, Masato; Maeda, Seiichiro
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 6 Pages, 2014/07
In order to develop MA contained MOX (MA-MOX) fuel design method, the analysis models to predict irradiation behavior of MA-MOX fuel have to be developed and the accuracy of irradiation behavior analysis code should be evaluated with the result of post-irradiation examinations (PIEs) for MA-MOX fuels. In this study, we developed the computer module "TRANSIT" to compute thermal properties of MA-MOX fuel. TRANSIT can give thermal conductivity, melting temperature and vapor pressures of MA-MOX. By using this module, we improved the thermal behavior analysis code "DIRAD" and developed DIRAD-TRANSIT code system to compute the irradiation behavior of MA-MOX fuel. This system was verified with the results of PIEs for the conventional MOX fuels and the MA-MOX fuels irradiated in the experimental fast reactor "JOYO". As the result of the verification, it can be mentioned that the DIRAD-TRANSIT system would precisely predict the fuel thermal behavior, i.e. fuel temperature and fuel restructuring, for oxide fuels containing several percent minor actinides.
Ito, Hiroto; Zheng, X.; Tamaki, Hitoshi; Maruyama, Yu
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 8 Pages, 2014/07
Takeda, Takeshi; Otsu, Iwao; Yonomoto, Taisuke
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 10 Pages, 2014/07
Yamano, Hidemasa; Tobita, Yoshiharu
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 10 Pages, 2014/07
In this paper, detailed descriptions are given for two representative experimental analyses (VECTORS and OMEGA), which are intended to validate high speed multi-phase flow dynamics in pin bundle structure and large vapor bubble expansion dynamics into a coolant pool, respectively. Through the experimental analyses, the SIMMER-III code has proved to be basically valid both numerically and physically, with currently applicability to integral reactor safety calculations.
Kofuji, Hirohide; Yano, Tetsuji*; Myochin, Munetaka; Matsuyama, Kanae*; Okita, Takeshi*; Miyamoto, Shinya*
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 6 Pages, 2014/07
For the research and development of the nuclear waste disposal concept suitable to the pyrochemical processing system and its performance evaluation, the iron-phosphate glass is examined as an alternative waste form for high level waste generated from electro-refining process. In order to enhance the waste element content in the glass matrix and improve the durability of the waste form, optimization experiments of the glass composition were carried out and the effect of additional other transition metal oxides was found out in this study.
Abe, Satoshi; Ishigaki, Masahiro; Shibamoto, Yasuteru; Yonomoto, Taisuke
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 9 Pages, 2014/07
Zheng, X.; Ito, Hiroto; Tamaki, Hitoshi; Maruyama, Yu
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 10 Pages, 2014/07
Integral severe accident code MELCOR 1.8.5 has been applied to estimating uncertainty of source term for the accident at Unit 2 of Fukushima Daiichi Nuclear Power Plant and to discussing important models or parameters influential to the source term. Forty-two parameters associated with models for the transportation of radioactive materials were chosen and narrowed down to 18 through a set of screening analysis. These 18 parameters in addition to 9 parameters relevant to in-vessel melt progression obtained by the preceded uncertainty study were inputted to the subsequent sensitivity analysis by Morris method. This one-factor-at-a-time approach can preliminarily identify inputs which have important effects on an output, and 17 important parameters were selected from the total of 27 parameters through this approach. The selected parameters have been integrated into uncertainty analysis by means of Latin Hypercube Sampling technique and Iman-Conover method, taking into account correlation between parameters. Cumulative distribution functions of representative source terms were obtained through the present uncertainty analysis assuming the failure of suppression chamber. Correlation coefficients between the outputs and uncertain input parameters have been calculated to identify parameters of great influences on source terms, which include parameters related to models on core components failure, models of aerosol dynamic process and pool scrubbing.
Ishikawa, Jun; Maruyama, Yu
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 6 Pages, 2014/07
Ito, Chikara; Naito, Hiroyuki; Oba, Hironori; Saeki, Morihisa; Ito, Keisuke; Ishikawa, Takashi; Nishimura, Akihiko; Wakaida, Ikuo; Sekine, Takashi
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 7 Pages, 2014/07
A high-radiation resistant optical fiber has been developed in order to investigate the interiors of the reactor pressure vessels and the primary containment vessels of the Fukushima Daiichi Nuclear Power Station. The radiation resistance of an optical fiber was improved by increasing the amount of hydroxyl up to 1000 ppm in pure silica fiber. The improved image fiber consists of common cladding and a large number of fiber cores made from pure silica that contains 1000 ppm hydroxyl. The transmissive rate of an infrared image was not affected after the irradiation of 1 MGy. We have developed the fiber-coupled LIBS system to detect plasma emission efficiently in near-infrared region. In addition, we have performed a 
ray dose rate measurement using an optical fiber of which scintillator is attached to the tip. As a result, the concept of applicability of a probing system using the high-radiation resistant optical fibers has been confirmed.
Kamiji, Yu; Noguchi, Hiroki; Terada, Atsuhiko; Yan, X.
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 5 Pages, 2014/07
HTGR produces not only electricity but high temperature heat for heat application systems. In the Middle East countries, large demand exists for cogeneration of water and electricity from desalination plant with power station. Desalination with an HTGR gas turbine system can efficiently meet this demand because such system can produce pure water from using only the waste heat. The waste heat of up to 248MWt is available for desalination. In this paper, heat and mass balance was calculated for a new concept of desalination system which is shown to increase use of waste heat by incrementing the number of thermal loading steps at the heat recovery section. The calculation was performed at steady water production condition to clarify the optimum steps of incremental loading. As a result, it was found that heat transfer area of heat recovery section in case of 3 BHs was 28% smaller than that of 2BHs.
Yamaguchi, Yoshihito; Katsuyama, Jinya; Udagawa, Makoto; Onizawa, Kunio; Nishiyama, Yutaka; Li, Y.*
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 8 Pages, 2014/07
Kamiyama, Kenji; Tobita, Yoshiharu; Suzuki, Toru; Matsuba, Kenichi
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 12 Pages, 2014/07
Cheng, S.; Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 7 Pages, 2014/07
Tagami, Hirotaka; Cheng, S.; Tobita, Yoshiharu; Guo, L.*; Zhang, B.*; Morita, Koji*
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 8 Pages, 2014/07
The object of this study is to develop new analytical methods to simulate unique phenomena in self-leveling behavior and implement it to SFR safety analysis code. The new methods are developed with assuming that the debris bed behaves as Bingham fluid from this feature. They are categorized into two main parts. The first part is particle interaction models to model the effect of particle-particle collisions. The second part is a large deformation method, which simulates Bingham fluid characteristic of debris bed. An experimental study of self-leveling behavior is analyzed to validate the new methods. The assessment results show that these methods provide a basis to develop analytical methods of self-leveling behavior of debris bed in the safety assessment of SFRs.
Kawada, Kenichi; Sato, Ikken; Tobita, Yoshiharu; Pfrang, W.*; Buffe, L.*; Dufour, E.*
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 10 Pages, 2014/07
Suzuki, Toru; Tobita, Yoshiharu; Kawada, Kenichi; Tagami, Hirotaka; Sogabe, Joji; Ito, Kei
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 10 Pages, 2014/07
Akabane, Masaaki*; Koizumi, Yasuo; Uchibori, Akihiro; Kamide, Hideki; Ohshima, Hiroyuki
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 6 Pages, 2014/07
A liquid droplet entrainment which appears under heat transfer tube failure accident in a steam generator of sodium-cooled fast reactors and causes wastage on an adjacent tube was examined in this study. The visualization experiment on a high-pressure air jet in a liquid pool was carried out. Filament-like ears and wisps pulled out from the wavy gas-liquid interface were observed. The ears and the wisps were broken off and entrained into the air jet. This process seems quite similar to the entrainment process in an annular dispersed flow in a pipe. The velocity of the entrained droplet was estimated from an image processing. The axial velocity of the entrained droplet increased as the air jet velocity increased. The Transversal directional velocity was much slower than the axial directional velocity. The data obtained from this experiment are very useful for the study of the heat transfer tube failure accident.
Horiguchi, Naoki; Yoshida, Hiroyuki; Kaneko, Akiko*; Abe, Yutaka*
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 6 Pages, 2014/07
no abstracts in English
