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Yamano, Hidemasa; Takano, Kazuya; Kurisaka, Kenichi; Kikuchi, Shin; Kondo, Toshiki; Umeda, Ryota; Sato, Rika; Shirakura, Shota*
Dai-28-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2024/06
This project studies investigation on safety design guideline and risk assessment technology for sodium-cooled fast reactor with the molten-salt heat storage system, development of evaluation method for heat transferring performance between sodium and molten-salt and improvement of the performance, and evaluation of chemical reaction characteristic between sodium and molten-salt and improvement of its safety. This paper describes the effect of sodium-molten salt heat transfer tube failure in addition to the project overview and progress.
Kikuchi, Shin; Sato, Rika; Kondo, Toshiki; Umeda, Ryota; Yamano, Hidemasa
Dai-28-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2024/06
no abstracts in English
Yamano, Hidemasa; Kurisaka, Kenichi; Takano, Kazuya; Kikuchi, Shin; Kondo, Toshiki; Umeda, Ryota; Shirakura, Shota*
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09
This project studies investigation on safety design guideline and risk assessment technology for sodium-cooled fast reactor with the molten-salt heat storage system, development of evaluation method for heat transferring performance between sodium and molten-salt and improvement of the performance, and evaluation of chemical reaction characteristic between sodium and molten-salt and improvement of its safety. The project overview is presented in this report.
Yamano, Hidemasa; Kurisaka, Kenichi; Takano, Kazuya; Kikuchi, Shin; Kondo, Toshiki; Umeda, Ryota; Shirakura, Shota*; Hayashi, Masaaki*
Proceedings of 8th International Conference on New Energy and Future Energy Systems (NEFES 2023) (Internet), p.27 - 34, 2023/00
Times Cited Count：0 Percentile：0.00(Green & Sustainable Science & Technology)This project studies investigation on safety design guideline and risk assessment technology for sodium-cooled fast reactor with the molten-salt heat storage system, development of evaluation method for heat transferring performance between sodium and molten-salt and improvement of the performance, and evaluation of chemical reaction characteristic between sodium and molten-salt and improvement of its safety. The project overview is presented in this report.
Umeda, Ryota; Kondo, Toshiki; Kikuchi, Shin; Kurihara, Akikazu
Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 9 Pages, 2021/08
In this study, in order to obtain the fundamental information on aerosol transport behavior between cells, the Multiple cells with Expandable connecting pipe Test facility (MET) was manufactured and preliminary experiments were performed. In the preliminary experiments, simulated particles were used in a test system with two cells connected horizontally or vertically, and their transport behavior was measured. As a result, it was possible to confirm the behavior of the simulated particles transporting to the horizontal or vertical cells from the results such as images and sedimentation data.
Umeda, Ryota; Shimoyama, Kazuhito; Kurihara, Akikazu
Nihon Genshiryoku Gakkai Wabun Rombunshi, 19(4), p.234 - 244, 2020/12
Sodium-water reaction caused by failure of the steam generator tube of sodium-cooled fast reactor induce the wastage phenomenon, which has erosive and corrosive feature. In this report, the authors have performed the self-wastage experiments under high sodium temperature condition to evaluate the effect of wastage form/geometry by using two types of initial defect such as the micro fine pinhole and fatigue crack, and water leak rate on self-wastage rate. Based on the consideration of crack type influence, it was confirmed that self-wastage rate did not strongly depend on the initial defect geometry. As a mechanism of the self-plug phenomenon, it is speculated that sodium oxide intervenes and inhibits the progress of self-wastage. The dependence of initial sodium temperature on self-wastage rate was clearly observed, and new self-wastage correlation was derived considering the initial sodium temperature.
Kurihara, Akikazu; Umeda, Ryota; Shimoyama, Kazuhito; Kikuchi, Shin
Nihon 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.
Umeda, Ryota; Shimoyama, Kazuhito; Kurihara, Akikazu
JAEA-Technology 2017-018, 70 Pages, 2017/08
JAEA-Technology-2017-018.pdf:9.67MB
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 Na
O). Based on the corrosive data, authors quantitatively evaluated the influence factors of wastage and formulated the average corrosive equations.
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.
Kurihara, Akikazu; Umeda, Ryota; Kikuchi, Shin; Shimoyama, Kazuhito; Ohshima, Hiroyuki
Nihon Genshiryoku Gakkai Wabun Rombunshi, 14(4), p.235 - 248, 2015/11
Sodium-water reaction would take place due to a breach of heat transfer tube in steam generator (SG) of sodium-cooled fast reactor (SFR), and the reaction jet may cause wear to the neighboring tubes by thermal and chemical effects, which is so-called target-wastage. Accordingly, failure propagation caused by target-wastage may potentially detract the secondary cooling system integrity. In previous study, a great number of target-wastage experiments have been carried out for candidate materials under practical SG operation conditions. Target-wastage rate was derived from macroscopic boundary factors of reaction jet. However, this mock-up approach is not versatile, and does not befit for large-scale SG design. Therefore, target-wastage should be focused for safety assessment of the various SG design. In this study, experiment apparatus and technique on composite oxidation type corrosion with flow (COCF), which is integral part of target-wastage, were constructed to figure out the separation effect of local wastage factors under the high temperature sodium hydroxide (NaOH) and sodium monoxide (NaO) environment mainly generated by SWR. The authors quantitatively evaluated the effect of material temperature and fluid velocity on COCF rate, and diffusion coefficient of Mod.9Cr-1Mo steel into NaOH-NaO. Besides, it was revealed that COCF was sodium-iron composite oxidation type corrosion from metallographic observation and element assay.
Kurihara, Akikazu; Umeda, Ryota; Shimoyama, Kazuhito; Abe, Yuta; Kikuchi, Shin; Ohshima, Hiroyuki
Nihon Kikai Gakkai Rombunshu, B, 79(808), p.2640 - 2644, 2013/12
Overheating tube rupture of adjacent tubes arises from water/steam leak in steam generators of sodium-cooled fast reactors. It is very important to predict the tube wall stress (tube wall temperature) with a high degree of accuracy on evaluation of overheating tube rupture, and is crucial to estimate quantitatively the heat transfer coefficient between reaction jet and adjacent tubes which is one of the major influencing factor. The authors carried out the sodium-water reaction test (SWAT-1R) under the simulated operation condition of a real plant, and measured the correlation between heat transfer coefficient and void fraction around an adjacent tube. The authors confirmed that thermal environment around an adjacent tube was inferable from measured data, and heat transfer correlation equation proposed by Hamada et al. was applicable to the operation condition at elevated pressure and temperature.
Beauchamp, F.*; Nishimura, Masahiro; Umeda, Ryota; Allou, A.*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 10 Pages, 2013/03
T91 is one of the material candidates of SGU tubes for future sodium-cooled fast reactors (SFRs). Wastage characterization of T91 is needed to evaluate the consequences for safety and the availability of the SGU. Six T91 target tubes were incorporated in the SWR test facility (SWAT-1R) of JAEA and subjected to reaction jets. All tubes were successfully penetrated by the reaction jets, and the wastage rates were determined. This paper describes the SWAT-1R facility, the test procedure and operating conditions, and brings out the main results and experience gained through the wastage experiments.
Kurihara, Akikazu; Umeda, Ryota; Yanagisawa, Hideki*; Ohshima, Hiroyuki
Dai-16-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.9 - 10, 2011/06
In the case of sodium-water reaction accident in a steam generator of sodium-cooled fast reactors (FRs), adjacent heat transfer tubes may be damaged due to high temperature environment of the reaction field. For the purpose of understanding the overheating tube rupture mechanism, an experimental study has been performed to clarify waterside heat transfer characteristics during up-flow in a vertical tube under the real plant part-load operation conditions in which safety margin is least. A test tube was heated rapidly and the time averaged heat flux was estimated using an inverse solution. It was conformed that the heat transfer on the wall changed from nucleate boiling to transient-film boiling all over the heating section and dried-out surface spread from downstream toward upstream. We improved the heat transfer correlations applied to RELAP5 code and made sure the adequacy of these correlations to evaluate tube overheating.
Kurihara, Akikazu; Ohshima, Hiroyuki; Shimoyama, Kazuhito; Umeda, Ryota
Nihon Kikai Gakkai Rombunshu, B, 77(776), p.964 - 968, 2011/04
Sodium reacts chemically with water in case of unexpected heat transfer tube failure in a steam generator (SG) of sodium-cooled fast breeder reactors (FBRs) and exothermic reaction produces reaction field with high temperature and high corrosive action (sodium-water reaction). Adjacent tubes are damaged due to erosive and corrosive environment of the reaction field (wastage). Therefore, it is integral to evaluate such sodium-water reaction phenomena with high accuracy for the safety assessment of FBRs. For the purpose of understanding the wastage mechanism, an experiment was carried out in which sodium hydroxide (NaOH) as the main reaction product collided with the tube material under high temperature conditions simulating the reaction field. We confirmed that the erosion-corrosion rate of tube material has a tendency to increase as the temperature and velocity of NaOH are raised.
Kurihara, Akikazu; Kikuchi, Shin; Umeda, Ryota; Shimoyama, Kazuhito; Ohshima, Hiroyuki; Narabayashi, Tadashi*
no journal, ,
The authors derived a new wastage correlation which is superposed by liquid droplet impinging erosion(LDI) and flow-accelarated corrosion(FAC) taken into account the local influencing factors for target-wastage under sodium hydroxide and sodium monoxide environment caused by sodium-water reaction. The authors report the applicability on new wastage correlation using the anamnestic target-wastage data.
Kurihara, Akikazu; Umeda, Ryota; Shimoyama, Kazuhito; Abe, Yuta; Kikuchi, Shin; Ohshima, Hiroyuki
no journal, ,
Overheating tube rupture of adjacent tubes arise from water/steam leak in steam generators of sodium- cooled fast reactors. It is very important to predict the tube wall stress (tube wall temperature) with a high degree of accuracy on evaluation of overheating tube rupture, and is crucial to estimate quantitatively the heat transfer coefficient between reaction jet and adjacent tubes which is one of the major influencing factor. The authors carried out the sodium-water reaction test (SWAT-1R) under the simulated operation condition of a real plant, and measured the correlation between heat transfer coefficient and void fraction around an adjacent tube. The authors confirmed that thermal environment around an adjacent tube was inferable from measured data, and heat transfer correlation equation proposed by Hamada et al. was applicable to the operation condition at elevated pressure and temperature.
Shimoyama, Kazuhito; Kurihara, Akikazu; Kikuchi, Shin; Umeda, Ryota; Ohshima, Hiroyuki
no journal, ,
Corrosion may occur on the tube surface due to chemical reaction between sodium and water (self-wastage) if water/steam leak proceed through the penetrating crack caused in the steam generator tube of sodium-cooled fast reactor. When the self-wastage goes along up to inside wall of tube, water leak rate will be larger and it will be likely to spread the affected area caused by sodium-water reaction. It is very important to clarify the self-wastage behavior for locally affected region and detection of the water leak in real plant. In this study, we performed the self-wastage experiments under high sodium temperature condition to evaluate the effect of wastage form/geometry and water leak rate on self-wastage rate in the pinhole type micro crack.
Kurihara, Akikazu; Umeda, Ryota; Shimoyama, Kazuhito; Abe, Yuta; Kikuchi, Shin; Ohshima, Hiroyuki
no journal, ,
The authors have carried out the sodium-water experiment to acquire the data for heat transfer coefficient between reaction jet and tube, for the purpose of clarifying the overheating tube rupture phenomena under water leak accident in steam generator of sodium-cooled fast reactor.
Kurihara, Akikazu; Kikuchi, Shin; Umeda, Ryota; Ohshima, Hiroyuki
no journal, ,
In order to clarify the wastage mechanism on the neighboring heat transfer tube in steam generator of sodium-cooled fast reactor, the impinging experiment was carried out under high-temperature sodium-hydroxide/sodium mono-oxide environment. We report the dependence of sodium-hydroxide/sodium mono-oxide temperature and collision velocity upon the thinning rate of material.
Kurihara, Akikazu; Umeda, Ryota; Shimoyama, Kazuhito; Kikuchi, Shin; Ohshima, Hiroyuki; Narabayashi, Tadashi*
no journal, ,
Wastage phenomena on adjacent tubes (target-wastage) arise from water/steam leak in steam generators of sodium-cooled fast reactors. Target-wastage is likely to be caused by liquid droplet impingement erosion (LDI) and flow-accelerated corrosion (FAC) in an environment marked by high-temperature and high-alkali (reaction jet) due to sodium-water reaction. The authors carried out flow-accelerated corrosion experiments as a part of phenomena clarification experiments for target-wastage by using tube material under high-temperature sodium-hydroxide and sodium monoxide conditions which are mainly generated by sodium-water reaction. New wastage correlations were derived from LDI and FAC data based on influencing factors which were formed on the periphery of an adjacent tube, and were confirmed those applicability to water leak event in this report.
