Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Ohshima, Hiroyuki
Nippon 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.
Kurihara, Akikazu; Umeda, Ryota; Shimoyama, Kazuhito; Kikuchi, Shin
Nippon Kikai Gakkai Rombunshu (Internet), 84(859), p.17-00382_1 - 17-00382_11, 2018/03
Wastage on adjacent tubes (target-wastage) arise from water/steam leak in steam generators of sodium-cooled fast reactors (sodium-water reaction). Target-wastage is likely to be caused by liquid droplet impingement erosion (LDI) and Na-Fe composite oxidation type corrosion with flow (COCF) in an environment marked by high temperature and high-alkali (reaction jet) due to sodium-water reaction. In the previous study, the authors quantitatively evaluated the effect of material temperature and fluid velocity on COCF rate, and revealed that COCF was sodium-iron composite oxidation type corrosion from metallographic observation and element assay. In this study, the applicability of new wastage correlations was confirmed for each tube in sodium-water reaction test with straight vertical tube bundle under practical steam generator operation condition. The authors established that the new wastage correlations were applicable to each tube of tube bundle in the above test, and the time progress of wastage was qualitatively investigated for the two penetrated tubes in the period including the water and/or steam blowdown.
Uchibori, Akihiro; Watanabe, Akira*; Takata, Takashi; Ohshima, Hiroyuki
Nippon Kikai Gakkai Rombunshu (Internet), 84(859), p.17-00394_1 - 17-00394_6, 2018/03
For assessment of the wastage environment under tube failure accident in a steam generator of sodium-cooled fast reactors, a mechanistic computer code called SERAPHIM calculating compressible multicomponent multiphase flow with sodium-water chemical reaction has been developed. The original SERAPHIM code is based on the finite difference method. In this study, unstructured mesh-based numerical method was developed and introduced into the SERAPHIM code to advance a numerical accuracy for the complex-shaped domain including multiple heat transfer tubes. Validity of the unstructured mesh-based SERAPHIM code was investigated through the analysis of an underexpanded jet experiment. The calculated pressure profile showed good agreement with the experimental data. Numerical analysis of water vapor discharging into liquid sodium was also performed. It was demonstrated that the proposed numerical method could be applicable to evaluation of the sodium-water reaction phenomenon.
Umeda, Ryota; Shimoyama, Kazuhito; Kurihara, Akikazu
JAEA-Technology 2017-018, 70 Pages, 2017/08
In case of the water leak into sodium in a SG of SFRs due to tube failure, reaction jet is formed by sodium-water reaction with exothermic heat. The reaction jet forms highly alkaline environment with high temperature and high pressure, which cause local thinning of adjacent heat transfer tubes (target wastage). In this report, for the purpose of elucidation of target wastage, the authors developed the experimental apparatus and experimental technique which enable the separate evaluation of wastage influence factors, including temperature, impingement velocity, reagent ratio and so on by using high temperature sodium hydroxide as major reaction product and sodium monoxide as secondary reaction product. In addition, the impingement corrosion experiments have been conducted by using high temperature reagents (NaOH and NaO). Based on the corrosive data, authors quantitatively evaluated the influence factors of wastage and formulated the average corrosive equations.
Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Kurihara, Akikazu; Hamada, Hirotsugu; Ohshima, Hiroyuki
JAEA-Research 2017-007, 61 Pages, 2017/07
For safety assessment of a steam generator of sodium-cooled fast reactors, it is necessary to evaluate the possibility of occurring tube failure propagation and of water leak rate under sodium-water reaction accident. In the previous studies, a computer code called LEAP-II calculating a wastage-type failure propagation and the water leak rate during long-time event progress was developed. In this study, a numerical method to evaluate the possibility of occurring overheating rupture was introduced into the LEAP-II code to expand application range of this code. The completed code is called LEAP-III. The test analysis on a tube bundle configuration demonstrated that the overheating rupture model could provide conservative prediction.
Umeda, Ryota; Kurihara, Akikazu; Shimoyama, Kazuhito
JAEA-Technology 2016-030, 50 Pages, 2016/12
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.
Akabane, Masaaki*; Horiki, Sachiyo*; Osakabe, Masahiro*; Koizumi, Yasuo; Uchibori, Akihiro; Ohno, Shuji; Ohshima, Hiroyuki
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05
Behavior of liquid droplets in a high-velocity gaseous jet was experimentally investigated to provide validation data for the evaluation method of sodium-water reaction phenomenon. The visualization experiment on the entrained liquid droplets in the air jet submerged in a water pool was carried out. Filament-like wisps from the wavy gas-liquid interface were observed. The wisps were broken off and entrained into the air jet. The velocity of the entrained liquid droplets was estimated from an image processing. The axial velocity of the liquid droplets increased as the air inlet velocity increased. Acceleration behavior of the liquid droplets was also confirmed quantitatively.
Uchibori, Akihiro; Ohshima, Hiroyuki
Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 6 Pages, 2014/11
A computer code called SERAPHIM calculating compressible multicomponent multiphase flow with sodium-water chemical reaction has been developed to evaluate wastage environment under tube failure accident in a steam generator of sodium-cooled fast reactors. In this study, the numerical model for liquid droplet entrainment and its transport was developed. The applicability of the model was investigated through the analysis of the basic experiment. It was demonstrated that our numerical model could reproduce the time to end of entrainment and the pressure variation during the occurrence of entrainment.
Jang, S.*; Takata, Takashi; Yamaguchi, Akira*; Uchibori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki
Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 8 Pages, 2014/11
Numerical quantification of the self-wastage phenomenon has been carried out using a multi-dimensional computational code: SERAPHIM. The width of the completely enlarged crack was investigated in this study. Several steps of numerical calculations were devised to reproduce transient self-wastage phenomenon caused by Sodium Water Reaction (SWR). In the analyses, 2-dimensional calculation was carried out to obtained thermal hydraulic properties in the reaction zone. The wastage amount was evaluated based on hypothetical Arrhenius equation by using the temperature and molar concentration of Sodium hydroxide. New analytical grid was created by exchanging the solid cells to fluid cells in the reaction based on the wastage amount evaluation. These series of procedure have been repeated. The width and the shape of the enlarged crack showed good agreement with the experimental results.
Saito, Kazuo*; Ishida, Toshihisa
JAERI-Tech 2001-039, 25 Pages, 2001/06
no abstracts in English
Okamoto, Koji*; *
JNC-TY9400 2000-016, 90 Pages, 2000/06
no abstracts in English
; Yoshida, Eiichi; Aoto, Kazumi
JNC-TN9400 2000-042, 112 Pages, 2000/03
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 700C 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
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.
Odano, Naoteru; Yamaji, Akio*; Ishida, Toshihisa
Journal of Nuclear Science and Technology, 37(Suppl.1), p.78 - 82, 2000/03
no abstracts in English
JNC-TN9410 2000-003, 52 Pages, 1999/12
In May, 1999, disassembly and cleansing of sodium residues contained in the large cold trap (50MWSG) were carried out. Two cold trap units, one from the primary sodium loop and the other from the for the secondary sodium loop were disassembled and cleaned. This report describes the procedures, methods, and tasks under taken in the clean-up effort, including countermeasures for safe handling of sodium. The disassembly of the cold trap was based an information regarding similar cleansing activities external to JNC. There was also same a priori knowledge of the type and amount of sodium-laden residues. As this result, we conducted disassembly and cleansing task as provisionally planned. In fact we learned that disassembly methods for the specific components could be conducted in an aerated atmosphere. We thus gained additional disassembly and sodium cleansing experience under manageable and safe conditions.
Hidaka, Akihide; Asaka, Hideaki; Ueno, Shingo*; Yoshino, T.*; Sugimoto, Jun
JAERI-Research 99-067, p.55 - 0, 1999/12
no abstracts in English
JNC-TN4400 99-002, 192 Pages, 1999/03
The tritium transport analysis code, TTT, has been validated using data from the low power test of Monju, and then its behaviour at along term full power operation of Monju in future has been estimated, when the estimated transport and distribution of tritium in the reactor system has been also compared with the result in Joyo and Phenix, which had been already experienced long term operations. The TTT code had been develpped using the tiritium and hydrogen transport model proposed by R. Kumar, ANL, and had been applied to the evaluation in Monju design work. After then, futhermore, the code has been improved using the data from long term operation of Joyo with MK-II core, and in this work the code has been validated for the first time for Monju data. The results from this work are as follows; (1)Comparison of the best fitted tritium source rates from cores in Joyo, Phenix and Monju makes an estimation of the major source from control rods, (2)The calculated tritium concentration in each medium for cooling and its change is a reasonable agreement to the measured, C/E=1.1, (3)The cover gas transport model cosidering isotopic exchange of H and H can reproduce reasonably the measured concentration distirbution of tritium in sodium and cover gas, (4)The tritium concentration in secondary sodium of Monju was about l/50 times as much as the primary one, which shows the acceraration effect on cold tarapping of tritium due to coprecipitation with permeated hydrogen through Evaporater (EV) heat conduction tube walls. The tritium cold trapping efficiency was estimated to be 1 for coprecipitation with hydrogen and 0.3 for isotopic exchange, respectively, (5)Tritium transport and distribution for along term full power operation of Monju in future was estimated, which could involve a excess factor to 4 at the maximum. The tritium concentration in sodium and Steam Generator (SG) water will be substantially saturated after somthing like 10 years full power operation, ...
; 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.
; *; *; *; 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.