Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Yamaguchi, Yoshihito; Mano, Akihiro; Li, Y.
Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 10 Pages, 2024/03
The steam generator (SG) is an important component of a pressurized water reactor. In addition, local wall-thinning has been reported in SG tubes. The burst differential pressure, considering both the internal and external pressures from the primary and secondary coolant systems, should be predicted for the failure probability evaluation or structural integrity assessment of SG tubes. In this study, based on the results of burst tests performed in Japan and the United States, we improved the existing burst pressure estimation method for SG tubes with wall-thinning. In addition, as an example of the utilization of the improved burst pressure estimation method, the conditional failure probabilities for SG tubes with local wall-thinning, which is necessary for probabilistic risk assessment and risk-informed decision making, are calculated considering the dimensions of the wall-thinning.
Yamaguchi, Yoshihito; Mano, Akihiro; Li, Y.
Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 10 Pages, 2022/07
The steam generator (SG) tube is one of the important components in pressurized water reactors. Flaws such as wall-thinning or stress corrosion cracking have been reported in SG tubes. The burst pressure where both the internal and external pressures from the primary and secondary coolant systems are considered must be predicted to assess the structural integrity of SG tubes. Burst tests were performed by various organizations. On the basis of the test results, failure estimation methods were proposed. In this study, previous burst test data and existing failure estimation methods for SG tubes with wall-thinning or crack were investigated. As a result, the coefficient of the existing estimation method for SG tube with uniform wall-thinning was updated. In addition, failure estimation methods that are suitable for SG tubes with crack or local wall-thinning were proposed by considering the effects of the flaw shape and size on the burst pressure. The applicability of the failure estimation methods was confirmed by comparing the predicted results with the burst test data in actual SG tubes.
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.
Oba, Toshihiro; Yanagihara, Takao; Kato, Chiaki; Hamada, Shozo
JAERI-Tech 2001-059, 36 Pages, 2001/09
The demonstration test for evaluating reliability of the acid recovery evaporator at Rokkasho Reprocessing Plant has been carried out at JAERI. For the nondestructive mesurement of the thickness of heat transfer tubes in the acid recovery evaporator and short tubes used in corrosion test, we have developed ultra sonic thickness measuring apparatus using immersion method with high resolution. This apparatus can measure and record tube thickness automatically with a personal computer. The results obtained by this apparatus are coincident with the results obtained by a destructive method using an optical microscope.
Saito, Kazuo*; Ishida, Toshihisa
JAERI-Tech 2001-039, 25 Pages, 2001/06
JAERI-Tech-2001-039.pdf:0.94MB
no abstracts in English
Maki, Akira; ; Taguchi, Katsuya; ; Shimizu, Ryo; Shoji, Kenji;
JNC TN8410 2001-012, 185 Pages, 2001/04
JNC-TN8410-2001-012.pdf:9.61MB
"The third technological meeting of Tokai Reprocessing plant (TRP)" was held in JNFL Rokkasyo site on March 14
, 2001. The technical meetings have been held in the past two times. The first one was about the present status and future plan of the TRP and second one was about safety evaluation work on the TRP. At this time, the meeting focussed on the corrosion experrience, in-service inspection technology and future maintenance plan. The report contains the proceedings, transparancies and questionnaires of the meeting are contained.
Doi, Masamitsu; Kiuchi, Kiyoshi; Yano, Masaya*; Sekiyama, Yoshio*
JAERI-Research 2001-020, 17 Pages, 2001/03
JAERI-Research-2001-020.pdf:0.7MB
no abstracts in English
; 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.
Shirakawa, Noriyuki*; *; *; *
JNC TJ9440 2000-008, 47 Pages, 2000/03
JNC-TJ9440-2000-008.pdf:1.96MB
The numerical thermohydraulic analysis of a LMFR component should involve its whole boundaly in order to evaluate the effect of chemical reaction within it. Therefore, it becomes difficult mainly due to computing time to adopt microscopic approach for the chemical reaction directly. Thus, the thermohydraulic code is required to model the chemically reactive fluid dynamics with constitutive correlations. The reaction rate denpends on the binary contact areas between components such as continuous liquids, droplets, solid particles, and bubbles. The contact areas change sharply according to the interface state between components. Since no experiments to study the jet flow with sodium-water chemical reaction have been done, the goal of this study is to obtain the knowledge of flow regimes and contact areas by analyzing the fluid dynamics of multi-pahse and reactive components mechanistically with the particle interaction method. For the first stage of the study, the applicability of this method to the nalysis of a liquid jet into the other liquid pool was investigated. Based on the literatures, we investigated the jet flow mechanisms and analyzed the experiment of a water jet into a gasoline pool. We also analyzed SWAT3/Run19 test, the jet flow in a rod bundle, to study the applicability of the method to a complicated boundary without a chemical reaction model. The calculated fluid dynamics was in good agreement with the experiment. Furthermore, we studied and formulated the paths of phase change and chemical reaction, and conceptually designed the adopting the heat-transfer-limited phase change model and the synthesizd reaction model with a water-hydrogen conversion ratio.
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
Shinozaki, Masayuki; *; Furusawa, Takayuki
JAERI-Tech 99-064, 46 Pages, 1999/08
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.
; ;
PNC TN9410 97-087, 142 Pages, 1997/07
Computer data analysis is planned as an essential process to facilitate and speed up the ISI of MONJU steam generator tubes using the ECT technique. This process compares the phase and amplitude of the signal in a vector window in order to identify and categories defects. The categorization of the inspection signal requires a high level of precision. The analysis test was carried out taking the best operational conditions for reference. From this, the most accurate classification conditions were established. The MONJU PSI signal data was used to check the effectiveness of the process. The results are as follows. (A) Verification of the set parameter for off line processing. Automatic classification is possible for almost all the support plate signals. Classification of all the weld and bend signals was not possible. Therefore, the set parameter was selected for the category in which there were the largest number of signals was established. (B) Verification of the analysis processing conditions. The established analysis conditions allow automatic classification for about 80 to 85% of the signal comparison factor cases. Furthermore, it is possible to classify all the signals by additional operator intervention. In this way it is possible to analysis and evaluate all the MONJU steam generator tube ISI data. (C) Improvement of the data base. Evaluation of MONJU PSI flaw detection data was carried out by set parameter analysis. FOllowing these results the necessary data base for ISI signal evaluation was created.
Futakawa, Masatoshi; *; Takada, Shoji;
Nuclear Technology, 118(1), p.83 - 88, 1997/04
Times Cited Count：1 Percentile：14.48(Nuclear Science & Technology)no abstracts in English
Watanabe, Katsutoshi; Shindo, Masami; Nakajima, Hajime; Koikegami, Hajime*; Higuchi, Makoto*; Nakanishi, Tsuneo*; Sahira, Kensho*; Marushichi, Koki*; Takeiri, Toshiki*; Saito, Teiichiro*; et al.
JAERI-Research 97-009, 62 Pages, 1997/02
JAERI-Research-97-009.pdf:4.82MB
no abstracts in English
; *; Ara, Katsuyuki
Electromagnetic Nondestructive Evalution, 0, p.223 - 230, 1997/00
no abstracts in English
Hamada, Hirotsugu; Tanabe, Hiromi
PNC TN9410 96-027, 41 Pages, 1995/12
If a sodium-water reaction jet was formed due to water leakage in an FBR steam generator(SG), neighboring tubes would suffer from overheating. On the safety aspect of the SGs, it is important to confirm that the neighboring tubes would not fail under such a severe overheating condition. So far, an analytical model using the structural integrity analysis code, FINAS, has been prepared and validated by the explosive torch overheating test data. This report presents the results on the overheating failure analysis of the under-sodium leak in the PFR superheater(SH), 1987. In the SH with slow steam dump system in 1987, neighboring overheated tubes are failed about 3 seconds after the SH isolation, which is shown both by the leak in the PFR and its analysis. For the SH in which a fast steam dump system was installed after the leak of 1987, the analysis shows no tube failure due to the fast steam depression and cooling effect inside. These results indicate that the FINAS model adequately predicts the overheating failure and the specific SH design and operation possibly result in further growth of the leak. It is concluded that steam blow effect is extremely important and the analysis model presented here is useful for the overheating failure evaluation of the SGs.
