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Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Ohshima, Hiroyuki
Nihon Kikai Gakkai Rombunshu (Internet), 86(883), p.19-00353_1 - 19-00353_6, 2020/03
Evaluation of occurrence possibility of tube failure propagation under sodium-water reaction accident is an important issue. In this study, a numerical analysis method to predict occurrence of failure propagation by overheating rupture was constructed to expand application range of an existing computer code. Applicability of the method was constructed through the numerical analysis of the experiment on water vapor discharging in liquid sodium.
Umeda, Ryota; Kurihara, Akikazu; Shimoyama, Kazuhito
JAEA-Technology 2016-030, 50 Pages, 2016/12
JAEA-Technology-2016-030.pdf:5.22MB
In case of tube failure of a steam generator in sodium-cooled fast reactors, the reaction jet with high temperature and high velocity under highly alkaline environment is formed by cited exothermic reaction (sodium-water reaction). When the high temperature reaction jet covers the adjacent tubes, the material strength of tube decreases in the high temperature condition, and the adjacent tube may be swollen and failed by inner pressure (overheating tube rupture). For evaluation of the overheating tube rupture, tube failure is judged by comparison the hoop stress loaded by inner pressure with stress strength standard defined as creep strength depending on tube temperature. Thus, it is important to confirm the validation of this failure criterion based on the findings obtained in the simulated experiment of overheating tube rupture. In this report, for consideration on the validation of the failure criteria and elucidation on the failure mode and strength characteristics of failure, the authors carried out the rapid heating rupture experiment for the thin single and double-walled 9Cr steel tubes at high temperature up to 1500 K by using TRUST-2 rig in the Japan Atomic Energy Agency.
; Yoshida, Eiichi; Aoto, Kazumi
JNC TN9400 2000-042, 112 Pages, 2000/03
JNC-TN9400-2000-042.pdf:8.55MB
A sodium-water reaction drove from the single tube break in steam generator of FBR might overheat labor tubes rapidly under internal pressure loadings. lf the temperature of tube wall becomes too high, it has to be evaluated that the stress of tube does not exceed the material strength limit to prevent the propagation of tube rupture. This study clarified the tensile and creep properties of Mod.9Cr-1Mo steel at ultra-high temperature which will be used in evaluation of the tube burst by sodium-water reaction. The strain rates for tensile test are from 10%/min to 10%/sec, and creep-rupture time is maximum 277sec. The range of test temperature is 700
C to 1300C. The main results obtained were as follows; (1)The evaluation data on the relationship between tensile strength and strain rate and creep-rupture strength in shorter time on Mod.9Cr-1Mo steel were acquired. (2)Short-term mechanical properties of Mod.9Cr-1Mo steel were evaluated based on the results of tensile and creep-rupture tests up to 1300C. As a result of the evaluation, recommended equation of creep-rupture strength in the short-term was proposed. (3)Tensile and creep-rupture strength of Mod.9Cr-1Mo steel tube showed the value which was higher than the 2 1/4Cr-1Mo steel, and it was proven to have the superior properties.
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
Hino, Ryutaro; *;
Nihon Genshiryoku Gakkai-Shi, 33(1), p.73 - 75, 1991/01
Times Cited Count：0 Percentile：0.3(Nuclear Science & Technology)no abstracts in English
*; *; *; Fujiki, Kazuo; Kosaka, Atsuo; *
Tech. Rep. Inst. At. Energy,Kyoto Univ., 211, p.1 - 57, 1989/05
no abstracts in English
*; Miyake, Osamu; Daigo, Yoshimichi; *
PNC TN941 82-100, 48 Pages, 1982/04
The Computer code LEAP II had been developed in order to analize failure propagation phenomena by the sodium-water reaction in the steam generator of LMFBR. Here reported is verification analysis of the LEAP code by using Runs 14 and 15 test results of Steam Generator Safety Test Facility (SWAT-3). The main results are as follows: (1)As the results of parametric survey, the effects of the significant parameters such as time mesh, jet division number, etc. to the code were understood. (2)In comparison with the test results of Runs 14 and 15 of SWAT-3, the LEAP code can estimate the phenomena conservatively enough.
Takano, Kazuya; Kurisaka, Kenichi; Yamano, Hidemasa
no journal, ,
As part of the development of risk assessment technologies for sodium-cooled fast reactor coupled to thermal energy storage system with sodium-molten salt heat exchanger, the database on the number of heat transfer tube failure in molten salt heat exchanger and exposure time of Concentrated Solar Power (CSP) system to molten-salt was surveyed on the basis of reported practices of incidents/accidents in the existing CSP systems with molten-salt heat storage system. Using the database, a risk assessment methodology of the heat transfer tube failure frequency was also studied with a Bayesian estimation method.
