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Takeda, Tetsuaki; Ichimiya, Koichi*; Yamauchi, Daiki*
Nihon Kikai Gakkai Rombunshu, B, 74(748), p.189 - 196, 2008/12
The objectives of this study are to clarify performances of a heat transfer enhancement method using porous material with high porosity and to discuss applicability of this method for a gas heating type steam reformer. A heat transfer experiment has been performed using a simulated steam reformer apparatus to obtain characteristics of heat transfer and pressure drop. From the results obtained in this experiment, this method showed a good performance regarding heat transfer. It was also found that an enhancement of heat transfer using porous material with high porosity is further improved under the high temperature conditions as compared with the other methods. There is the possibility of the adoption of this method to the steam reformer of the hydrogen production system connected to a high temperature gas-cooled reactor.
Takeda, Tetsuaki*; Ichimiya, Koichi*; Nishio, Hitoshi*; Nakagawa, Shigeaki; Takamatsu, Kuniyoshi
Proceedings of 7th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-7) (CD-ROM), 11 Pages, 2008/10
Safety demonstration tests using the HTTR (High Temperature Engineering Test Reactor) are being performed to verify the inherent safety features and to validate the numerical code for the safety assessment of the VHTR (Very High Temperature Reactor). The partial loss of coolant flow test was performed under the condition of the ATWS (Anticipated Transient without Scram). We are planning to perform the test of loss of coolant flow and stopping the vessel cooling system (VCS). The test of the loss of coolant flow as one of safety demonstration tests is carried out by tripping all gas circulators, and the stopping VCS test is performed continuously after the loss of coolant flow. The objective of this study is to evaluate the temperature distribution of the reactor pressure vessel (RPV) and the VCS during the tests. It is necessary to consider the effect of thermal radiation from the RPV for evaluation of temperature of the VCS and concrete vessel.
Takeda, Tetsuaki; Ichimiya, Koichi*
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 4 Pages, 2007/04
As for the development of the coupling technology between the HTGR and the hydrogen production system, JAEA have carried out the hydrogen production test with the steam reforming process by natural gas. In the HTGR hydrogen production system, disk type fins are attached on the outside surface of the catalyst tube and the tube is inserted into the guide tube to increase an amount of transferred heat in the present design of the steam reformer. However, we have to take the deterioration of the structure strength by attaching the fins and processing the tube surface into consideration with the design of the steam reformer. The objectives of this study are to develop a method for heat transfer enhancement using a porous material and to discuss the applicability of this method into the steam reformer of the nuclear hydrogen production system. An experiment has been performed using the simulated apparatus of the steam reformer to obtain the heat transfer and fluid flow characteristics.
Ichimiya, Koichi*; Takeda, Tetsuaki; Uemura, Takuya*; Norikuni, Tetsuya*
Nihon Kikai Gakkai Rombunshu, B, 72(723), p.2747 - 2752, 2006/11
no abstracts in English
Takase, Kazuyuki; Kitamura, Tatsuaki*; Kume, Etsuo; Ichimiya, Koichi*; Komada, Ichiro*
Nihon Kikai Gakkai Kanto Shibu Yamanashi Koenkai (2003) Koen Rombunshu, No.030-4, p.77 - 78, 2003/00
no abstracts in English
Uemura, Takuya*; Takeda, Tetsuaki; Ichimiya, Koichi*
Nihon Kikai Gakkai Yamanashi Koenkai Koen Rombunshu (020-4), p.49 - 50, 2002/10
A heat transfer experiment was performed using a horizontal circular tube to obtain the heat transfer and fluid flow characteristics in the tube inserted copper wire. From the results obtained in this experiment, it was found that an amount of heat removal in the tube with copper wire inserts increased about 20% comparing with a tube having a smooth wall. A heat transfer coefficient of the tube inserted copper wire also increased 30% to 50% under the constant pumping power condition.
; Kunugi, Tomoaki;
Album of Visualization,No. 13, 0, p.19 - 20, 1996/00
no abstracts in English
Akino, Norio; Kunugi, Tomoaki; Shiina, Yasuaki; ;
Flow Visualization,V, p.87 - 92, 1990/00
no abstracts in English
Takeda, Tetsuaki; Ichimiya, Koichi*; Yamauchi, Daiki*
no journal, ,
A technology development of a hydrogen production system by a nuclear heat is being performed as a heat application system of a high-temperature gas cooled reactor in the Japan Atomic Energy Research Institute. The objective of this study is to clarify heat transfer characteristics of the steam reformer inserted metallic wire with high porosity. An experiment has been performed using a double coaxial channel and single channel with metallic wire to obtain the heat transfer characteristics. From the results obtained in this experiment, it seems to be probable that an enhancement of heat transfer method using metallic wire inserts is further improved under the high temperature conditions in efficiency as compared with the conventional reformer.
Ichimiya, Koichi*; Takeda, Tetsuaki; Yamauchi, Daiki*
no journal, ,
A hydrogen production system using nuclear heat is performed as a heat application system of the HTTR in JAERI. In the present studies heat transfer improvement in the reformer tube is examined using a porous material with high porosity. In the first report, the heat transfer and flow characteristics of an annular tube were examined experimentally to obtain the basic data. These results were compared with the convenient data.
Takeda, Tetsuaki; Nishio, Hitoshi*; Ichimiya, Koichi*
no journal, ,
Safety demonstration tests using the HTTR (High Temperature Engineering Test Reactor) are being performed to verify the inherent safety features and to validate the numerical code for the safety assessment of the VHTR (Very High Temperature Reactor). The test of loss of coolant flow as one of safety demonstration tests is carried out by tripping of the helium gas circulators. The objective of this study is to evaluate the temperature distribution of the reactor pressure vessel (RPV) and the vessel cooling system (VCS) during the loss of coolant flow. The temperature distribution of the RPV and surrounding concrete structure were obtained using a commercially available analysis code STAR-CD. The effect of thermal radiation from the RPV was evaluated using the analytical and experimental results. It was found that the RPV and surrounding concrete structure were cooled enough by the VCS in the case of loss of coolant flow.
Takeda, Tetsuaki; Tochio, Daisuke; Inaba, Yoshitomo; Ichimiya, Koichi*; Nishio, Hitoshi*
no journal, ,
Safety demonstration tests using the HTTR (High Temperature Engineering Test Reactor) are being performed to verify the inherent safety features and to validate the numerical code for the safety assessment of the VHTR (Very High Temperature Reactor). The test of loss of coolant flow as one of safety demonstration tests is carried out by tripping of the helium gas circulators. The objective of this study is to evaluate the temperature distribution of the reactor pressure vessel (RPV) and the vessel cooling system (VCS) during the loss of coolant flow. The temperature distribution of the RPV and surrounding concrete structure were obtained using a commercially available analysis code STAR-CD. The effect of thermal radiation from the RPV was evaluated using the analytical and experimental results. It was found that it is important to take account of the effect of thermal radiation in the transient analysis to evaluate the temperature change of the concrete accurately.
Takeda, Tetsuaki; Nishio, Hitoshi*; Ichimiya, Koichi*
no journal, ,
Safety demonstration tests using the HTTR (High Temperature Engineering Test Reactor) are being performed to verify the inherent safety features and to validate the numerical code for the safety assessment of the VHTR (Very High Temperature Reactor). The test of loss of coolant flow as one of safety demonstration tests is carried out by tripping of the helium gas circulators. The objective of this study is to evaluate the temperature distribution of the reactor pressure vessel (RPV) and the vessel cooling system (VCS) during the loss of coolant flow. The temperature distribution of the RPV and surrounding concrete structure were obtained using a commercially available analysis code STAR-CD. It is necessary to consider the effect of thermal radiation from the RPV for evaluation of temperature of the VCS and concrete vessel.
