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Tamai, Hidesada; Onuki, Akira; Kureta, Masatoshi*; Liu, W.; Sato, Takashi; Akimoto, Hajime
Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), 8 Pages, 2005/05
no abstracts in English
Tamai, Hidesada; Kureta, Masatoshi; Yoshida, Hiroyuki; Akimoto, Hajime
JSME International Journal, Series B, 47(2), p.293 - 298, 2004/05
no abstracts in English
Ishiyama, Shintaro; Muto, Yasushi; Fumizawa, Motoo; Tsukazaki, Kazuo*
Nihon Genshiryoku Gakkai Wabun Rombunshi, 1(2), p.164 - 176, 2002/06
no abstracts in English
Islam, M. S.*; Terada, Atsuhiko*; Kinoshita, Hidetaka; Hino, Ryutaro; Monde, Masanori*
JAERI-Tech 2001-044, 49 Pages, 2001/07
JAERI-Tech-2001-044.pdf:2.49MB
no abstracts in English
Takeda, Tetsuaki
JAERI-Research 2000-056, 27 Pages, 2000/12
JAERI-Research-2000-056.pdf:1.86MB
no abstracts in English
M.S.Islam*; Hino, Ryutaro; Haga, Katsuhiro; Monde, Masanori*; Sudo, Yukio
Journal of Nuclear Science and Technology, 35(9), p.671 - 678, 1998/09
Times Cited Count：17 Percentile：77.77(Nuclear Science & Technology)no abstracts in English
; Kamide, Hideki;
PNC TN9410 98-083, 118 Pages, 1998/07
Large-scaled thermohydraulic tests are planned for new key technologies in the heat transport systems of a demonstration fast reactor. The test facility is consisted of components from a reactor vessel to a steam generator (SG). Basic design of the large-scaled thermohydraulic test facility is 1/3 scale of the demonstration fast reactor with two primary cooling loops and two into one secondary loop. The secondary piping length of the test facility is longer than the 1/3 scale of the demonstration fast reactor. The tests facility has the branch and junction of the secondary piping because of two primary loops and one SG. There is a possibility of flow and temperature unbalance if a buoyancy force were large and pressure loss were small. Therefore, dynamics analyses of the thermal transition tests had been done in which the secondary piping length. To examine the unbalance occurred or not, the natural circulation analysis had been performed providing different heat transfer area of the IHX or presser loss of the primary loop between A loop and B loop. It was shown from the analyses that the temperature response during the transition was delayed in the test model compared to the real reactor. Main cause of the delay was due to the real scaled SG. Other parameters, the length of piping etc., were not very influential to the response. The analysis such predicted that there wasn't large difference of global behaviors between the loops. Therefore, it was shown that there would be no problem, if the difference were made between the loops due to a manufacturing error.
; Ohshima, Hiroyuki
PNC TN9410 97-104, 69 Pages, 1997/12
As the part of the improvement of single-phase subchannel analysis code ASFRE-III, verification study about fuel-pin heat transfer model and flow resistance model of the code was carried out. Temperature distributions in a fuel pin predicted by the fuel-pin heat transfer model of ASFRE-III were compared with those calculated by the structure analysis code FINAS, which has been well validated and applied to various structure analyses, using the same boundary conditions. The comparison showed that the results by these two codes agreed with maximum difference of 1 %. and therefore the validity of the model was confirmed. With respect to the flow resistance model, distributed resistance model (DRM), which can enhance the consistent description of the fluid flow and wire-spacer interaction, was examined through analyses of two hydraulic simulation tests using the fifth mock-up fuel subassembly for the prototype LMFBR and the second mock-up fuel subassembly for the experimental rector. The calculated pressure difference between pressure measurement taps whose positions were near the top and the bottom of the fuel-pin bundle agreed with the measured data of both tests. The predicted pressure distribution in a horizontal cross section was also compared with the calculational result by the finite element analysis code SPIRAL and agreement was good.
PNC TJ1409 97-021, 27 Pages, 1997/03
1 Reflection of R&D Results, Design and Operation Experiences All knowledge obtain in the project, such as R&D results, design, operation experiences and so on, are tbe reflected to following items. 1) Improvement of safety and reliability in plant eration 2) Design modification of the plant 3) Design of the next plant 2 Basic Sndpoint for Data Base Composition "Design/R&D/Plant Performance Data Base" will be mposed ad shown in the following, so that the Data Base could be utilized for reflecng them to the above items and for improving the above items, effectively and efficitly. (1) United Data Bases of Design and R&D "Design/R&D/Plant Performance Data Ba" will be composed by uniting design data base and R&D data base, considering that t R&D of the project is mainly made in order to establish the design engineering and chnical basis, such as design policy, design criteria, design conditions, alloable dign limits, design verification, etc. (2) Addition of Initial Plant Performa
Ohshima, Hiroyuki
PNC TN9410 96-128, 82 Pages, 1996/04
The purpose of the study is to offer an analytical tool for investigations of local subassembly accidents which have been recognized as a key issue of the safety assessment of liquid metal cooled fast breeder reactors due to the dense structure of the fuel grid and the high power density. The porous blockage model, which was implemented in the single-phase subchannel analysis code ASFRE-III last year, was upgraded for the accuracy improvement. Correlation equations on pressure loss, heat conduction and heat transfer based on the packed bed theory was newly applied to the model in order to more precisely evaluate thermal hydraulic phenomena in the porous blockage regions. The model verification analysis was carried out on the basis of the data of porous blockage simulation test Scarlet-2 performed in France. In the analysis, it was confirmed that this model could reproduce coolant temperature distribution in the blockage region and could predict the peak temperature with high accuracy. The detailed model validation will be carried out by using the water/sodium out-of-pile tests which have scheduled to be conducted at PNC in the near future.
Shimomura, Hiroaki
Nihon Genshiryoku Gakkai-Shi, 38(10), p.803 - 813, 1996/00
Times Cited Count：1 Percentile：14.44(Nuclear Science & Technology)no abstracts in English
; Hino, Ryutaro; ; Nakahira, Masataka; *; Furuya, Kazuyuki; Tada, Eisuke; Seki, Masahiro
Fusion Engineering and Design, 28, p.103 - 112, 1995/00
Times Cited Count：1 Percentile：17.53(Nuclear Science & Technology)no abstracts in English
Kurasawa, Toshimasa; Takatsu, Hideyuki; Sato, Satoshi; Mori, Seiji*; Hashimoto, T.*; Nakahira, Masataka; Furuya, Kazuyuki; Tsunematsu, Toshihide; Seki, Masahiro; Kawamura, Hiroshi; et al.
Fusion Engineering and Design, 27, p.449 - 456, 1995/00
Times Cited Count：7 Percentile：59.17(Nuclear Science & Technology)no abstracts in English
*; *
PNC TJ9381 94-001, 84 Pages, 1994/02
The study in last year examined the realization of 1000MWt natural circulation ATR. In this study. The sensitivity analysis is did making for the design. The results are as below. (1)The influence of riser pipe diameter. The natural circulation on riser pipe 3B is analyzed. 3B pipe is cheaper than 5B riser pipe that is a base condition. The results are that fuel dryout is happen by 1.8MW channel power on 3B, but is not happen on 5B riser pipe. (2)The influence of the riser pipe number reduction and the riser pipe diameter change. The case of connection 2 pressure tubes by Y piece pipe is analyzed. In this case, the riser pipe number is half. The cases that riser pipe is 5B and 4B are analyzed. The results are as below. (a)Natural circulation flow lower by riser pipe number is half. (b)Fuel dryout is happen by 1.8MW channel power on 4B, but is not happen on 5B. (3)The influence of the middle position header connection and the riser pipe diameter change, The case of connection 50 pressure tubes by the middle position header is analysed. The cases that riser pipe is 24B and 32B are analyzed. The results are as below. (a)In 2 cases, natural circulation flow lower as compared with base case. (b)In 2 cases, fuel dryout are not happen by 1.8MW channel power. (4)The influence of pressure tube diameter. The case of pressure tube diameter widening is analyzed. So, the pressure loss of pressure tube domain govern the flow. (a)Natural circulation flow increase by pressure tube diameter widening. (b)The flow swing is by pressure tube diameter widening. (c)In analysis cases, fuel dryout is not happen. We get 1.8MW channel power on the condition that riser pipe is 4B, riser pipe is connected by Y piece pipe, pressure tube diameter is 122.0mm.
Ishitsuka, Etsuo; Nakamichi, Masaru; Kawamura, Hiroshi; Sagawa, Hisashi; *; *; Saito, Minoru
Journal of Nuclear Materials, 212-215, p.881 - 884, 1994/00
Times Cited Count：2 Percentile：28.09(Materials Science, Multidisciplinary)no abstracts in English
*; Okamoto, Koji*; *; Fumizawa, Motoo
Nihon Kikai Gakkai Rombunshu, B, 59(566), p.172 - 178, 1993/10
no abstracts in English
Ishitsuka, Etsuo; Kawamura, Hiroshi; Sagawa, Hisashi; Nagakura, Masaaki*; Kanzawa, Toru*
JAERI-M 93-044, 24 Pages, 1993/03
no abstracts in English
