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Herranz, L. E.*; Jacquemain, D.*; Nitheanandan, T.*; Sandberg, N.*; Barr
, F.*; Bechta, S.*; Choi, K.-Y.*; D'Auria, F.*; Lee, R.*; Nakamura, Hideo
Progress in Nuclear Energy, 127, p.103432_1 - 103432_14, 2020/09
Times Cited Count：4 Percentile：16.23(Nuclear Science & Technology)Oka, Hiroshi; Tanno, Takashi; Otsuka, Satoshi; Yano, Yasuhide; Kaito, Takeji
Nuclear Materials and Energy (Internet), 16, p.230 - 237, 2018/08
Times Cited Count：4 Percentile：38.11(Nuclear Science & Technology)Oka, Hiroshi; Tanno, Takashi; Otsuka, Satoshi; Yano, Yasuhide; Uwaba, Tomoyuki; Kaito, Takeji; Onuma, Masato*
Nuclear Materials and Energy (Internet), 9, p.346 - 352, 2016/12
Times Cited Count：21 Percentile：88.55(Nuclear Science & Technology)Nakajima, Norihiro; Nishida, Akemi; Miyamura, Hiroko; Iigaki, Kazuhiko; Sawa, Kazuhiro
Kashika Joho Gakkai-Shi (USB Flash Drive), 36(Suppl.2), 4 Pages, 2016/10
Since nuclear power plants have dimensions approximately 100m
and their structures are an assembly made up of over 10 million components, it is not convenient to experimentally analyze its behavior under strong loads of earthquakes, due to the complexity and hugeness of plants. The proposed system performs numerical simulations to evaluate the behaviors of an assembly like a nuclear facility. The paper discusses how to carry out visual analysis for assembly such as nuclear power plants. In a result discussion, a numerical experiment was carried out with a numerical model of High Temperature engineering Test Reactor of Japan Atomic Energy Agency and its result was compared with observed data. A good corresponding among them was obtained as a structural analysis of an assembly by using visualization. As a conclusion, a visual analytics methodology for assembly is discussed.
Nakajima, Norihiro; Nishida, Akemi; Kawakami, Yoshiaki; Suzuki, Yoshio; Sawa, Kazuhiro; Iigaki, Kazuhiko
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05
A numerical analysis controlling and managing system is implemented on K, which controls the modelling process and data treating, although the manager only controls a structural analysis by finite element method. The modeling process is described by the list of function ID and its procedures in a data base. The manager executes the process by order in the list for simulation procedures. The manager controls the intention of an analysis by changing the analytical process one to another. Experiments were carried out with static and dynamic analyses.
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.
Takahashi, Yoshikazu; Yoshida, Kiyoshi; Nabara, Yoshihiro; Edaya, Masahiro*; Mitchell, N.*
IEEE Transactions on Applied Superconductivity, 16(2), p.783 - 786, 2006/06
Times Cited Count：9 Percentile：46.59(Engineering, Electrical & Electronic)To investigate the conductor behavior during a quench, quench tests of Center Solenoid (CS) insert coils were carried out with various initial conditions in DC and pulse modes. The conductor has very similar configuration and parameters. The inductive heater, attached at the center of the length, initiated an artificial quench in DC mode. A quench has also occurred during the pulse operation with the ramping rate of 0.4-2 T/s. Simulations of electric, thermal and hydraulic behaviors of the conductor during the quench tests in both modes were carried out by using the thermohydraulic simulation code. The experimental results were compared with the simulation and good agreement was obtained. These results are described and the implication for quench detection in ITER is discussed in this paper. The voltage tap method will be used for the quench detection for the CS, and the sensitivity of the detection and the maximum temperature of the conductor during a quench are described. It is shown that the detection system could be designed with high enough detection sensitivity.
Okumura, Keisuke; Kawasaki, Kenji*; Mori, Takamasa
JAERI-Research 2005-018, 64 Pages, 2005/08
JAERI-Research-2005-018.pdf:3.26MB
In the KRITZ-2 critical experiments, criticality and pin power distributions were measured at room temperature and high temperature (about 245 degree C) for three different cores loading slightly enriched UO
or MOX fuels. For nuclear data testing, benchmark analysis was carried out with a continuous-energy Monte Carlo code MVP and its four nuclear data libraries based on JENDL-3.2, JENDL-3.3, JEF-2.2 and ENDF/B-VI.8. As a result, fairly good agreements with the experimental data were obtained with any libraries for the pin power distributions. However, the JENDL-3.3 and ENDF/B-VI.8 give under-prediction of criticality and too negative isothermal temperature coefficients for slightly enriched UO cores, while the older nuclear data JENDL-3.2 and JEF-2.2 give rather good agreements with the experimental data. From the detailed study with an infinite unit cell model, it was found that the differences among the libraries are mainly due to the different fission cross section of U-235 in the energy rage below 1.0 eV.
Sumita, Junya; Shibata, Taiju; Nakagawa, Shigeaki; Hanawa, Satoshi; Iyoku, Tatsuo; Ishihara, Masahiro
Transactions of 18th International Conference on Structural Mechanics in Reactor Technology (SMiRT-18), p.4822 - 4828, 2005/08
Graphite materials are used for structural components in High Temperature Gas-Cooled Reactor (HTGR) core because of their excellent thermo/mechanical properties. Thermal conductivity of graphite components is reduced by neutron irradiation in reactor operation. The reduced conductivity is expected to be recovered by thermal annealing effect when irradiated graphite component is heated above irradiated temperature. In the present study, temperature analyses considering the annealing effect of the HTGR core at a depressurization accident were carried out and influence of annealing effect on maximum fuel temperature was investigated. The analyses show that the annealing effect can reduce the fuel temperature about 100
C at the maximum, and it is possible to evaluate the maximum fuel temperature more appropriately. It was also shown that the core-temperature of High Temperature Engineering Test Reactor (HTTR) at the safety demonstration tests can be analyzed with the developed evaluation method considering annealing effect.
Inaba, Yoshitomo; Zhang, Y.*; Takeda, Tetsuaki; Shiina, Yasuaki
Heat Transfer-Asian Research, 34(5), p.293 - 308, 2005/07
Water cooling panels have been adopted as the vessel cooling system of the HTTR to cool the reactor core indirectly by natural convection and thermal radiation. In order to investigate the heat transfer characteristics of high temperature gas in a vertical annular space between the reactor pressure vessel and cooling panels of the HTTR, we carried out experiments and numerical analyses on natural convection heat transfer coupled with thermal radiation heat transfer in an annulus between two vertical concentric cylinders with the inner cylinder heated and the outer cylinder cooled. In the present experiments, Rayleigh number based on the height of the annulus ranged from 2.0
10
to 5.410 for helium gas and from 1.210
to 3.510 for nitrogen gas. The numerical results were in good agreement with the experimental ones regarding the surface temperatures of the heating and cooling walls. As a result of the experiments and the numerical analyses, the heat transfer coefficient of natural convection coupled with thermal radiation was obtained.
Uchida, Shunsuke*; Sato, Tomonori; Morishima, Yusuke*; Hirose, Tatsuya*; Miyazawa, Takahiro*; Kakinuma, Nagao*; Sato, Yoshiyuki*; Usui, Naoshi*; Wada, Yoichi*
Proceedings of 12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors (CD-ROM), p.19 - 29, 2005/00
Static and dynamic responses of stainless steel specimens exposed to HO and O in high temperature water were evaluated by analyzing ECP and FDCI (frequency dependent complex impedance). The oxide films on the specimens were characterized by multilateral surface analyses, e.g., LRS, SIMS, XPS and direct electric resistance measurement. As a result of evaluation, it was confirmed that (1) corrosive condition of BWR normal water chemistry (NWC) was simulated by 100 ppb HO without co-existing O, while that of hydrogen water chemistry (HWC) was simulated by 10 ppb HO, (2) ECP under HWC was as high as that under NWC, while dissolution rate of oxide film under HWC was much lower than that under NWC, (3) combination effects of electric resistance and dissolution rate of oxide caused same level ECP for both NWC and HWC, and (4) distinct weight loss of the specimen exposed to 100 ppb HO was observed.
Tachibana, Yukio; Sawahata, Hiroaki; Iyoku, Tatsuo; Nakazawa, Toshio
Nuclear Engineering and Design, 233(1-3), p.89 - 101, 2004/10
Times Cited Count：10 Percentile：55.63(Nuclear Science & Technology)no abstracts in English
Takamatsu, Kuniyoshi; Furusawa, Takayuki; Hamamoto, Shimpei; Nakagawa, Shigeaki
Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 11 Pages, 2004/10
Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are in progress to verify the inherent safety features for High Temperature Gas-cooled Reactors (HTGRs). The coolant flow reduction test by tripping one or two out of three gas circulators is one of the safety demonstration tests. The reactor power safely brings to a stable level without a reactor scram and the temperature transient of the reactor-core is very slow. Through the safety demonstration test, the two dimensional temperature analysis code (TAC-NC code) was improved. This paper describes the validation of the TAC-NC code using the measured value of the test by tripping of one and two out of three gas circulators at 30%(9MW). The TAC-NC code could evaluate accurately the temperature transient within 10% during the test. Also, it was confirmed that the temperature transient by tripping all gas circulators is very slow.
Sumita, Junya; Nakano, Masaaki*; Tsuji, Nobumasa*; Shibata, Taiju; Ishihara, Masahiro
JAERI-Tech 2004-055, 25 Pages, 2004/08
JAERI-Tech-2004-055.pdf:4.25MB
Neutron irradiation remarkably reduces the thermal conductivity of graphite, and the reduced thermal conductivity is recovered by annealing effect if the graphite is heated above the irradiation temperature. Therefore, it is expected that the reduced thermal conductivity of graphite components in the HTGR could be recovered by the annealing effect in accidents, such as a depressurization accident. Then, an analytical investigation of the annealing effect on thermal performance of a HTGR core was carried. The analysis showed that the annealing effect reduces the maximum fuel temperature about 70C, and it is important to introduce the annealing effect appropriately in the temperature analysis of the core components and reactor internals. In addition, an annealing test method was investigated to evaluate the effect quantitatively, and the test plan was made.
Inaba, Yoshitomo; Zhang, Y.*; Takeda, Tetsuaki; Shiina, Yasuaki
Nihon Kikai Gakkai Rombunshu, B, 70(694), p.1518 - 1525, 2004/06
no abstracts in English
Inoue, Masaki*; Iwai, Takashi; Arai, Yasuo; Asaga, Takeo*
Proceedings of GLOBAL2003 Atoms for Prosperity; Updating Eisenhower's Global Vision for Nuclear Energy (CD-ROM), p.1694 - 1703, 2003/11
no abstracts in English
Takada, Eiji*; Fujimoto, Nozomu; Matsuda, Atsuko*; Nakagawa, Shigeaki
JAERI-Tech 2003-040, 23 Pages, 2003/03
JAERI-Tech-2003-040.pdf:1.28MB
In the High Temperature Engineering Test Reactor (HTTR), since the primary circuit is very high at the high temperature test operation, the special alloy Alloy800H is used as the metallic material for cladding tubes and spines of the control rods to endure the temperature of 950 degrees centigrade. The control rod is supposed to be exchanged for the excess use of its temperature limit 900 degrees centigrade according to the strength data of Alloy800H. The scram shutdown by loss of off-site electric power at the high temperature test operation is assumed as an event of the temperature of the control rods to exceed 900 degrees centigrade. In this report, the temperature of the control rods is analyzed by using the measurement data of the rise-to-power test. The result of this analysis it is confirmed that the control rod temperature does not exceed its limitation value even after the most temperature raises event of the loss of off-site electric power at the high temperature test operation.
Ishihara, Masahiro; Baba, Shinichi; Takahashi, Tsuneo*; Aihara, Jun; Shibata, Taiju; Hoshiya, Taiji
JAERI-Tech 2002-054, 169 Pages, 2002/07
JAERI-Tech-2002-054.pdf:5.93MB
no abstracts in English
Nagase, Fumihisa; Otomo, Takashi; Uetsuka, Hiroshi
JAERI-Research 2001-009, 21 Pages, 2001/03
JAERI-Research-2001-009.pdf:2.93MB
no abstracts in English
Okumura, Susumu; Kurashima, Satoshi; Ishimoto, Takayuki*; Yokota, Wataru; Arakawa, Kazuo; Fukuda, Mitsuhiro; Nakamura, Yoshiteru; Ishibori, Ikuo; Nara, Takayuki; Agematsu, Takashi; et al.
Proceedings of 13th Symposium on Accelerator Science and Technology, p.283 - 285, 2001/00
no abstracts in English
