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Kobayashi, Jun; Aizawa, Kosuke; Ezure, Toshiki; Nagasawa, Kazuyoshi*; Kurihara, Akikazu; Tanaka, Masaaki
JAEA-Research 2022-009, 125 Pages, 2023/01
JAEA-Research-2022-009.pdf:29.22MB
The design studies of an advanced loop-type sodium-cooled fast reactor (Advanced- SFR) have been carried out by the Japan Atomic Energy Agency (JAEA). At the core outlet, temperature fluctuations occur due to mixing of hot sodium from the fuel assembly with cold sodium from the control rod channels and radial blanket assembly. These temperature fluctuations may cause high cycle thermal fatigue around a bottom of Upper Internal Structure (UIS) located above the core. Therefore, we conducted a water experiment using a 1/3 scale 60 degree sector model that simulated the upper plenum of the advanced loop-type sodium-cooled reactor. And we proposed some countermeasures against large temperature fluctuations that occur at the bottom of the UIS. In this report, we have summarized that the effect of the countermeasure structure to mitigate the temperature fluctuation generated at the bottom of UIS is confirmed, and the Reynolds number dependency of the countermeasure structure and the characteristics of the temperature fluctuation on the control rod surface.
Kobayashi, Jun; Aizawa, Kosuke; Ezure, Toshiki; Kurihara, Akikazu; Tanaka, Masaaki
Hozengaku, 20(3), p.89 - 96, 2021/10
Hot sodium from the fuel assembly can mix with cold sodium from the control rod (CR) channel and the blanket assemblies at the bottom plate of the Upper Internal Structure (UIS) of Advanced-SFR. Temperature fluctuation due to mixing of the fluids at different temperature between the core outlet and cold channel may cause high cycle thermal fatigue on the structure around the bottom of UIS. A water experiment using a 1/3 scale 60 degree sector model simulating the upper plenum of the Advanced-SFR has been conducted to examine countermeasures for the significant temperature fluctuation generated around the bottom of UIS. We focused on the temperature fluctuations near the primary and backup control rod channels, and studied the countermeasure structure to mitigate the temperature fluctuation through temperature distribution and flow velocity distribution measurements. As a result, effectiveness of the countermeasure to mitigate the temperature fluctuation intensity was confirmed.
Kobayashi, Jun; Aizawa, Kosuke; Ezure, Toshiki; Kurihara, Akikazu; Tanaka, Masaaki
Hozengaku, 20(3), p.97 - 101, 2021/10
Focusing on the thermal striping phenomena that occurs at a bottom of the internal structure of an advanced sodium-cooled fast reactor (Advanced-SFR) that has been designed by the Japan Atomic Energy Agency, a water experiment using a 1/3 scale 60 degree sector model simulating the upper plenum of the Advanced-SFR has been conducted to examine countermeasures for the significant temperature fluctuation generated around the bottom of Upper Internal Structure (UIS). In the previous paper, we reported the effect of measures to mitigate temperature fluctuations around the control rod channels. In this paper, the same test section was used, and a water experiment was conducted to obtain the characteristics of temperature fluctuations around the radial blanket fuel assembly. And the shape of the Core Instrumentation Support Plate (CIP) was modified, and it was confirmed that it was highly effective in alleviating temperature fluctuations around the radial blanket fuel assembly.
Akita, Koichi; Shibahara, Masakazu*; Ikushima, Kazuki*; Nishikawa, Satoru*; Furukawa, Takashi*; Suzuki, Hiroshi; Harjo, S.; Kawasaki, Takuro; Vladimir, L.*
Yosetsu Gakkai Rombunshu (Internet), 35(2), p.112s - 116s, 2017/06
Abe, Hitoshi; Masaki, Tomoo; Amano, Yuki; Uchiyama, Gunzo
JAEA-Research 2014-022, 12 Pages, 2014/11
JAEA-Research-2014-022.pdf:1.03MB
To contribute safety evaluation of boiling and drying accident of high active liquid waste (HALW) in fuel reprocessing plant, release behavior of Ru, which was considered as an important nuclide for evaluating public dose from the volatile viewpoint, has been investigated. It has been reported that release of Ru becomes conspicuously after HALW is dried up. In this work, to grasp the release behavior of Ru, release ratio of Ru with thermal decomposition of Ru nitrate, which would be in the dried HALW, was measured and release rate constant of Ru from the nitrate was estimated. It was found that the calculation result of release rate of Ru from the nitrate with rise of temperature by using the constant could well simulate the result acquired from the beaker-scale experiment.
Kaneko, Tetsuji; Tsukatani, Ichiro; Kiuchi, Kiyoshi
JAERI-Tech 2004-035, 18 Pages, 2004/03
JAERI-Tech-2004-035.pdf:0.81MB
Fuel elements used in the Reduced-Moderation Water Reactor (RMWR) have the stacking structure consisting of MOX pellets and UO
blankets in a fuel rod in order to attain the high breeding ratio and high burn-up simultaneously. It is a characteristic of the fuel elements that there is high thermal stress caused by inhomogeneous linear power density along the longitudinal direction of the fuel rod in comparison with the present LWR fuels. For this reason, it is important to estimate local deformation behavior of the fuel cladding tube with temperature difference caused by MOX pellet and UO blanket. The testing machine was designed to investigate thermal-fatigue behavior under biaxial stress condition. The testing machine consists of the temperature distribution control unit, low cycle fatigue testing unit and internal pressure loading unit, it is also possible to conduct the simulation tests to investigate effects of pressure change with burn-up and longitudinal load change due to operation modes and restriction of fuel rods.
Kasahara, Seiji; Hwang, G.*; Nakajima, Hayato; Choi, H.*; Onuki, Kaoru; Nomura, Mikihiro
Journal of Chemical Engineering of Japan, 36(7), p.887 - 899, 2003/07
Times Cited Count：70 Percentile：88.24(Engineering, Chemical)Thermal efficiency of the IS thermochemical hydrogen production process was evaluated. Sensitivities of operation conditions (HI conversion ratio, pressure and reflux ratio at HI distillation and concentration of HI after EED) and nonidealities of the process (electric energy loss in EED, loss at heat exchangers and loss of waste heat recovery as electricity) were investigated. Concentration of HI after EED had the most significant effect of 13.3 % on thermal efficiency in operation conditions. Nonidealities had importance on thermal efficiency. Thermal efficiency was 56.8 % with optimized operation conditions and no nonidealities.
Ashikagaya, Yoshinobu; Yoshino, Toshiaki; Yasu, Katsuji; Kurosawa, Yoshiaki; Sawa, Kazuhiro
JAERI-Tech 2002-094, 80 Pages, 2002/12
JAERI-Tech-2002-094.pdf:12.8MB
no abstracts in English
Ezato, Koichiro; Dairaku, Masayuki; Taniguchi, Masaki; Sato, Kazuyoshi; Akiba, Masato
Journal of Nuclear Materials, 307-311(Part1), p.144 - 148, 2002/12
Times Cited Count：14 Percentile：64.59(Materials Science, Multidisciplinary)no abstracts in English
Yanagi, Yoshihiko*; Kosaku, Yasuo; Hatano, Toshihisa; Kuroda, Toshimasa*; Enoeda, Mikio; Akiba, Masato
JAERI-Tech 2002-046, 45 Pages, 2002/05
JAERI-Tech-2002-046.pdf:2.61MB
no abstracts in English
Kurosawa, Masayoshi; Naito, Yoshitaka; Suyama, Kenya; ; Suzuki, Katsuo*;
Nihon Genshiryoku Gakkai-Shi, 40(6), p.486 - 494, 1998/00
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)no abstracts in English
Shimomura, Hiroaki
JAERI-Research 96-034, 73 Pages, 1996/06
JAERI-Research-96-034.pdf:2.47MB
no abstracts in English
Suzuki, Satoshi; Akiba, Masato; Araki, Masanori; Sato, Kazuyoshi; Yokoyama, Kenji; Dairaku, Masayuki
Journal of Nuclear Materials, 212-215(1), p.1365 - 1369, 1994/09
no abstracts in English
; Akiba, Masato; Araki, Masanori; Dairaku, Masayuki; Ise, Hideo*;
JAERI-M 91-085, 20 Pages, 1991/05
no abstracts in English
; Akiba, Masato; Araki, Masanori; Seki, Masahiro; Ise, Hideo*; ; ; ; Yamazaki, Seiichiro*
Proc. of the 2nd Japan Int. SAMPE Symp. on Advanced Materials for Future Industries,Needs and Seeds, p.1176 - 1182, 1991/00
no abstracts in English
JAERI-M 85-204, 7 Pages, 1985/12
no abstracts in English
; Koizumi, Yasuo; Tasaka, Kanji
Nucl.Eng.Des., 86, p.219 - 239, 1985/00
Times Cited Count：2 Percentile：37.22(Nuclear Science & Technology)no abstracts in English
; ; ;
Kyowa Giho, 309, p.2195 - 2197, 1983/00
no abstracts in English
Tone, Tatsuzo
Journal of Nuclear Science and Technology, 16(6), p.453 - 456, 1979/00
Times Cited Count：1no abstracts in English
Shindo, Masami; ; Kondo, Tatsuo
Proc.of 2nd Japan-US HTGR Safety Technology Seminar,Material Properties and Design Method Session, 11 Pages, 1978/00
no abstracts in English
; ; ;
JAERI-M 6799, 37 Pages, 1976/11
no abstracts in English
