Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Matsushita, Kentaro; Ezure, Toshiki; Imai, Yasutomo*; Fujisaki, Tatsuya*; Tanaka, Masaaki
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2021-Nendo Koen Rombunshu (Internet), 4 Pages, 2021/08
For evaluation of gas entrainment phenomenon at free surface in reactor vessel of sodium-cooled fast reactor, the gas entrainment evaluation tool named "Stream Viewer" has been developed. In Stream Viewer, depth of surface vortex dimple is predicted by calculating pressure decrease at the vortex center using velocity distribution around the vortex and Burgers vortex model. In this report, a method to identify continuous vortex center lines from a velocity distribution is newly developed. It becomes possible to evaluate three-dimensional distribution of pressure decrease along vortex center line. Then, the method is validated by applying Stream Viewer to an open channel experiment. As the result, it was confirmed that vortex center lines were successfully identified by the improved Stream Viewer. Moreover, it was also shown that the evaluation accuracy of gas entrainment was expected to be improved by considering distribution of pressure decrease along vortex center line.
Ueno, Fumiyoshi; Irisawa, Eriko; Kato, Chiaki; Igarashi, Takahiro; Yamamoto, Masahiro; Abe, Hitoshi
Proceedings of European Corrosion Congress 2016 (EUROCORR 2016) (USB Flash Drive), 7 Pages, 2016/09
In this study, we focused on the effect of the boiling of nitric acid solution on the corrosion of a stainless steel-made concentrator in reduced pressure in fuel reprocessing plant. In order to perform the simulation test in a non-radioactive condition, nitric acid solution with the addition of vanadium as an oxidizing metal ion were used. Corrosion tests were carried out under the conditions of boiling at reduced pressure, and of non-boiling at normal pressure and several temperatures. As a result, corrosion was accelerated by the solution boiling while it was not by non-boiling at the same temperature. It was found also that the temperature dependence of corrosion rate is the same in the both conditions of boiling and non-boiling. The corrosion accelerating effect will be discussed on the basis of the reaction among nitric acid, NOx and vanadium, etc.
Irisawa, Eriko; Ueno, Fumiyoshi; Kato, Chiaki; Abe, Hitoshi
Zairyo To Kankyo, 65(4), p.134 - 137, 2016/04
In order to investigate the effect of boiling under reduced pressure on corrosion of stainless steel in the nitric acid solution, the corrosion tests simulating the high-level radioactive liquid waste evaporator were performed. The results of immersion tests of stainless steels in the solution with and without boiling showed that the corrosion rates in boiling solution were larger than those in not boiling solution in case of same temperature of solution. Moreover, the cathode polarization curves showed that the corrosion potential of stainless steel in boiling solutions were shifted nobler, and the current intensity became larger than that in not boiling solutions. According to these results, it can be concluded that boiling of solution under reduced pressure accelerate the corrosion rates.
Otani, Hiroshi; Mizui, Hiroyuki; Higashiura, Norikazu; Bando, Fumio*; Endo, Nobuyuki*; Yamagishi, Ryuichiro*; Kume, Kyo*
Heisei-25-Nendo Koeki Zaidan Hojin Wakasawan Enerugi Kenkyu Senta kenkyu Nempo, 16, P. 66, 2014/10
no abstracts in English
Takamatsu, Kuniyoshi; Nakagawa, Shigeaki
Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(4), p.369 - 380, 2004/12
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). By upgrading the code, the analytical result can evaluate accurately the measured value of the transient temperature distribution within 20
C. Also, the improved code can analyze the maximum fuel temperature and temperature distributions of the test by tripping all the three gas circulators. The result of this study can be applied to development for not only the commercial HTGRs but also the Very High Temperature Reactor (VHTR) such as one of the Generation IV reactors.
Katanishi, Shoji; Kunitomi, Kazuhiko; Tsuji, Nobumasa*; Maekawa, Isamu*
Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(3), p.257 - 267, 2004/09
no abstracts in English
Katanishi, Shoji; Kunitomi, Kazuhiko
Nihon Genshiryoku Gakkai Wabun Rombunshi, 2(1), p.55 - 67, 2003/01
no abstracts in English
O
coating as electrical insulatorNakamichi, Masaru; Kawamura, Hiroshi
Fusion Engineering and Design, 58-59, p.719 - 723, 2001/11
Times Cited Count：11 Percentile：61.99(Nuclear Science & Technology)no abstracts in English
Antariksawan, A. R.*; Hidaka, Akihide; Moriyama, Kiyofumi; Hashimoto, Kazuichiro*
JAERI-Tech 2001-011, 116 Pages, 2001/03
JAERI-Tech-2001-011.pdf:4.02MB
no abstracts in English
Kojima, Kazutoshi; Yoshikawa, Masahito; Oshima, Takeshi; Ito, Hisayoshi; Okada, Sohei
Materials Science Forum, 338-342, p.1239 - 1242, 2000/00
no abstracts in English
Hidaka, Akihide; Asaka, Hideaki; Ueno, Shingo*; Yoshino, T.*; Sugimoto, Jun
JAERI-Research 99-067, p.55 - 0, 1999/12
JAERI-Research-99-067.pdf:2.51MB
no abstracts in English
Hamada, Hirotsugu; *; *; *; Hiroi, Hiroshi*
PNC TN9410 98-029, 122 Pages, 1998/05
The following items have been studies to evaluate overheating failure of FBR generator heat transfer tubes: (1)To establish a structural integrity analysis method. The strength standard values for 2.25Cr-1Mo steel was established taking account of time dependent effect to overheating failure mechanism based on high temperature (700 - 1200C) creep data and was validated by tube rupture simulation test data. (2)To improve and validate blow down analytical method. The analytical result by use of BLOOPH, the FBR blow down code, was compared with that by use of RELAP-5, the general purpose thermo-hydraulic code, and a good agreement was obtained. (3)To quantitatively validate the entire overheating analysis model by sodium water reaction data Sodium-water reaction tests of SWAT-3 and LLTR were analyzed using above mentioned analytical method. The ductile fracture occurred earlier than the creep fracture in the analysis and the comparison of tube failure times with the experiments showed sufficient conservativeness. Based on the above studies, the analytical method was applied to PFR superheater leak event and the Monju steam generator accidental analysis. The followings were quantatitively shown through the analysis: (1)The most important cause that multi-tube failure occurred in the 1987 PFR superheater-2 leak is that the superheater did not equip a fast steam dump system at the time of the leak event. (2)Overheating failure will not occur under any operational conditions of Monju in both steady state and transient phases such as water/steam blow-down. (3)Although safety margin becomes small when the water/steam flow rate becomes small during the blow-down, the modification of the plant such as hastening blow-down by equipping more relief valves will drastically improve the safety margin.
Takase, Kazuyuki; Kunugi, Tomoaki; Seki, Yasushi
Eighth Int. Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), 3, p.1321 - 1327, 1997/00
no abstracts in English
Enoeda, Mikio; Kawamura, Yoshinori; Okuno, Kenji
Fusion Technology, 30(3), p.885 - 889, 1996/12
no abstracts in English
Hidaka, Akihide; Ezzidi, A.*; Sugimoto, Jun
PSA 96: Int. Topical Meeting on Probabilistic Safety Assessment, 3, p.1548 - 1556, 1996/00
no abstracts in English
Watanabe, Norio; *; *; Kumamaru, Hiroshige; Hirano, Masashi
The 3rd JSME/ASME Joint Int. Conf. on Nuclear Engineering, Vol. 3, 0, p.1303 - 1310, 1995/00
no abstracts in English
Nemoto, Takeshi; Okada, Takashi; ; Ouchi, Jin; Kondo, Isao
PNC TN8410 92-119, 45 Pages, 1992/06
None
*; Kiuchi, Kiyoshi; *;
Proc. of the Int. Symp. on Material Chemistry in Nuclear Environment, p.469 - 477, 1992/00
no abstracts in English
Hidaka, Akihide; Sugimoto, Jun; *; Soda, Kunihisa
JAERI-M 91-175, 65 Pages, 1991/10
no abstracts in English
Nemoto, Takeshi; Ouchi, Jin; Okada, Takashi; ; Kondo, Isao;
PNC TN8410 91-014, 31 Pages, 1991/01
None
