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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.
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, 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
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
Garcia Rodriguez, D.; Matsubara, Shinichiro
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 9 Pages, 2014/07
Structural reliability of the circumferentially cracked core support mount of Monju FBR is analyzed using FEA. The 3D shell model employed was derived after detailed evaluation of the core support mount behavior with a specific 3D solid model. 1st, elastoplastic static analysis shows that, under nominal operating conditions, the overall structure would be able to survive a total loss of the core support mount. 2nd, using the double elastic slope method it was inferred that earthquake loading integrity could be warranted up to a crack representing over 50% of the total circumference.
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.
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
Abe, Satoshi; Ishigaki, Masahiro; Shibamoto, Yasuteru; Yonomoto, Taisuke
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 9 Pages, 2014/07
Tanaka, Masaaki; Miyake, Yasuhiro*
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 13 Pages, 2014/07
In this study, numerical simulation for the WATLON experiment which was the water experiment of a T-junction piping system (T-pipe) was carried out to validate the MUGTHES and to investigate the relation between the mechanism of temperature fluctuation generation and the unsteady motion of large eddy structures. In the numerical simulation, the large eddy simulation (LES) approach with standard Smagorinsky model was employed as eddy viscosity model to simulate large scale eddy motion in the T-pipe. As for uncertainty quantification in the validation process, the modified method of the Grid Convergence Index (GCI) estimation based on the least squire version could successfully quantify uncertainty. Through the numerical simulations, it was indicated that the fine mesh arrangement could improve the temperature distribution in the wake. It could be found that the thermal mixing phenomena in the T-pipe were caused by the mutual interaction of the necklace-shaped vortex around the wake from the front of the branch jet, the horseshoe-shaped vortex and the Karman's vortex motions in the wake.
Nishimura, Akihiko; Terada, Takaya; Takenaka, Yusuke*; Furuyama, Takehiro*; Shimomura, Takuya
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 6 Pages, 2014/07
Since 2007, JAEA has been developing laser based technologies of structural health monitoring. The FBG sensor made by femtosecond laser processing will be the best candidate. To make the best use of the heat resistant characteristic, the FBG sensor was embedded in metal mold by laser cladding. A groove was processed to the surface of a SUS metal plate. We used a QCW laser to weld a filler wire on the plate. A series of weld beads perfectly formed a sealing clad on the groove. Though the FBG sensor was buried tightly, no degradation on the reflection spectrum was detected after the processing. The FBG sensor could detect the vibration of the plate caused by impact shocks and audio vibration. The reflection peak of the FBG sensor under laser cladding condition was shifted to be 6 nm. We demonstrated that the corresponded temperature derive from the reflection peak shift reached 600 degrees in heat shock experiments. The installation procedure of a FBG sensor using a portable laser cladding machine was described.
Isono, Kenichi; Kubo, Shigenobu; Chikazawa, Yoshitaka; Dozaki, Koji*; Oya, Takeaki*; Yui, Masahiro*
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 7 Pages, 2014/07
Aiming at enabling maintenance and repair of almost all components in JSFR demonstration reactor to a level equivalent to that attained for the light water reactors, we identified a number of parts which have difficulty in maintenance and repair in main components of the reactor structure and the primary/secondary main coolant system. And we defined the criteria for design improvement and then provided candidates of improvement measures for the identified parts. Furthermore, we made a modification of the plant design in a consistent manner integrating the improvements investigated for each major component. A series of evaluations were conducted to check the feasibility as a power plant. As the result, we found that the concept could be adopted not only to the demonstration reactor (750 MWe) but to the commercial one (1500 MWe).
Takaya, Shigeru
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 7 Pages, 2014/07
This paper presents evaluation methods of fatigue strength of similar and dissimilar welded joints of modified 9Cr-1Mo steel which is a candidate structural material for a demonstration fast breeder reactor being developed in Japan. The discontinuity of mechanical properties across welded joint causes a non-homogeneous strain distribution, and this effect should be taken into account for evaluation of fatigue strength of weld joints. In this study, 2-element model, which is consisted of base metal and weld metal, was employed. Firstly, strain ranges of each element are calculated, and secondly fatigue lives of each element are evaluated. Finally, shorter fatigue life is chosen as fatigue life of the weld joint. Failure position can be also estimated by this model. Evaluation results were compared with experimental data at elevated temperature, and it was shown that they agree well.
Takaya, Shigeru; Machida, Hideo*; Kamishima, Yoshio*
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 4 Pages, 2014/07
This paper describes the outline of the guidelines on structural reliability evaluation for the passive components of the fast breeder reactor (FBR). The guidelines are now being prepared by the task force for the system based code in the Japan society of mechanical engineers in order to contribute to reducing differences in evaluated structural reliability by evaluators. They consist of five chapters, which are "General rules", "Reliability evaluation", "Failure scenario setting", "Modeling", and "Failure probability calculation", respectively. Details of each chapter are explained.
Koizumi, Yasuo*; Ote, Naosuke*; Kamide, Hideki; Ohno, Shuji; Ito, Kei
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 7 Pages, 2014/07
