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Journal Articles

Study on muliti-dimensional core cooling behavior of sodium-cooled fast reactors under DRACS operating conditions

Ezure, Toshiki; Onojima, Takamitsu; Tanaka, Masaaki; Kobayashi, Jun; Kurihara, Akikazu; Kameyama, Yuri*

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.3355 - 3363, 2019/08

Steady-state sodium experiments under the operating conditions of a decay heat removal system (DHRS) were carried out as part of the safety enhancement of sodium-cooled fast reactors using the PLANDTL 2 facility, which has 30 heated channels with electric heaters and 25 no-heated channels as the simulated core. In the experiments, a direct reactor auxiliary cooling system (DRACS) with a dipped type direct heat exchanger (DHX) in the upper plenum was used as the DHRS. This paper reports on the preliminary experimental results of the PLANDTL 2 experiments under the DRACS operating conditions without flow in the primary circuit, including the thermal hydraulic interactions between the upper plenum and the core under the DHX operating conditions and the resulting core cooling behavior from the outside of the multiple rows of the fuel assemblies

JAEA Reports

Material test data of SUS316 and SUS321, 1

Hashidate, Ryuta; Kato, Shoichi; Kurihara, Akikazu

JAEA-Data/Code 2019-005, 117 Pages, 2019/07

JAEA-Data-Code-2019-005.pdf:2.54MB

SUS316 and SUS321 are used for structural materials of the Fast Breeder Reactors (FBRs), because of excellent high creep strength. This report summarized the mechanical properties data on SUS316 and SUS321 obtained in various tests including the long-term material tests and the material tests in sodium.

Journal Articles

Improvement of steam generator tube failure propagation analysis code LEAP for evaluation of overheating rupture

Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Kurihara, Akikazu; Hamada, Hirotsugu; Ohshima, Hiroyuki

Journal of Nuclear Science and Technology, 56(2), p.201 - 209, 2019/02

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

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 developed to expand application range of an existing computer code. Applicability of the method was demonstrated through the numerical analysis of the experiment on water vapor discharging in liquid sodium.

Journal Articles

Measurement of Velocity Field in Five Jets Water Test (FIWAT) for thermal striping in sodium-cooled fast reactor

Aizawa, Kosuke; Kobayashi, Jun; Tanaka, Masaaki; Kurihara, Akikazu; Ishida, Katsuji*; Nagasawa, Kazuyoshi*

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 10 Pages, 2018/11

A conceptual design of an advanced loop type sodium cooled reactor has been carried out in the Japan Atomic Energy Agency (JAEA). Temperature fluctuation is caused by mixing of fluids at different temperature from the control rod channels and the core fuel assemblies, high cycle thermal fatigue may arise on the Core Instrument Plane (CIP) at bottom of the Upper Internal Structure (UIS). In JAEA, 1/3-scaled five jets water tests (FIWAT) have been performed in order to investigate thermal striping phenomena around the CIP. In this study, the velocity field was measured in the mixing area between the jet outlet and the bottom of the structure by using particle image velocimetry (PIV) to compare with the temperature fluctuation characteristics.

Journal Articles

Evaluation of target-wastage in consideration of sodium-water reaction environment formed on the periphery of an adjacent tube in steam generator of sodium-cooled fast reactor

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.

JAEA Reports

Phenomenon elucidation experiment for target wastage caused in steam generator of sodium-cooled fast reactor; Corrosion experiment in flowing high-temperature sodium hydroxide environment

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$$_{2}$$O). Based on the corrosive data, authors quantitatively evaluated the influence factors of wastage and formulated the average corrosive equations.

JAEA Reports

Development of LEAP-III code for evaluation of long-time event progress under tube failure accident in steam generators

Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Kurihara, Akikazu; Hamada, Hirotsugu; Ohshima, Hiroyuki

JAEA-Research 2017-007, 61 Pages, 2017/07

JAEA-Research-2017-007.pdf:4.3MB

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.

Journal Articles

Study on reactor vessel coolability of sodium-cooled fast reactor under severe accident condition; Water experiments using a scale model

Ono, Ayako; Kurihara, Akikazu; Tanaka, Masaaki; Ohshima, Hiroyuki; Kamide, Hideki; Miyake, Yasuhiro*; Ito, Masami*; Nakane, Shigeru*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04

The water experiment apparatus simulating the thermal hydraulics in a reactor vessel under operating the decay heat removal systems (DHRSs) was fabricated. The theoretical evaluation for similarity and results of basic experiments show applicability for a scale model experiment of a sodium-cooled fast reactor. This paper, moreover, describes the results of flow visualization experiment under operating a dipped-type passive DHX, which is planned to be installed in both a loop type reactor and pool type reactor, and the calculation results using FLUENT comparing with the result of water experiment.

Journal Articles

Reaction path and product analysis of sodium-water chemical reactions using laser diagnostics

Deguchi, Yoshihiro*; Muranaka, Ryota*; Kamimoto, Takahiro*; Takagi, Taku*; Kikuchi, Shin; Kurihara, Akikazu

Applied Thermal Engineering, 114, p.1319 - 1324, 2017/03

The purpose of this study aims to clarify the gas phase sodium-water reaction path and reaction products quantitatively. The counter-flow diffusion experiment device was employed to analyze the reaction path and reaction products using laser diagnostics. The main product of sodium-water reaction was determined to be NaOH and the sodium oxide was not notably measured compared with NaOH.

JAEA Reports

Rapid heating rupture experiment using the high chromium steel tubes

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.

Journal Articles

Experimental measurement of vortex cavitation around a suction pipe inlet

Ezure, Toshiki; Ito, Kei; Kameyama, Yuri*; Kurihara, Akikazu; Kunugi, Tomoaki*

Konsoryu, 30(2), p.189 - 196, 2016/06

Journal Articles

Development of a multiphysics analysis system for sodium-water reaction phenomena in steam generators of sodium-cooled fast reactors

Uchibori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki

AIP Conference Proceedings 1702, p.040010_1 - 040010_4, 2015/12

 Times Cited Count:0 Percentile:100

A Multiphysics analysis system was newly developed to evaluate possibility of failure propagation occurrence in a steam generator of sodium-cooled fast reactors. The system consists of the computer codes, SERAPHIM, TACT, and RELAP5. The SERAPHIM code calculates the multicomponent multiphase flow involving sodium-water chemical reaction. Applicability of the SERAPHIM code was confirmed through the analyses of the basic experiment and the experiment on water vapor discharging in liquid sodium. The TACT code was developed to calculate heat transfer from the reacting jet to the adjacent tube and to predict the tube failure occurrence. The numerical models of the TACT code were verified through some related experiments. The RELAP5 code evaluates thermal hydraulic behavior of water inside the tube. The original heat transfer correlations were corrected for the rapidly heated tube. The developed system enabled us to evaluate the wastage environment and possibility of failure propagation.

Journal Articles

Study on target wastage for sodium-water reaction environment formed on periphery of adjacent tube in steam generator of sodium-cooled fast reactor; Composite oxidation-type corrosion with flow experiment using high-temperature sodium hydroxide

Kurihara, Akikazu; Umeda, Ryota; Kikuchi, Shin; Shimoyama, Kazuhito; Ohshima, Hiroyuki

Nippon 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 (Na$$_{2}$$O) 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-Na$$_{2}$$O. Besides, it was revealed that COCF was sodium-iron composite oxidation type corrosion from metallographic observation and element assay.

Journal Articles

Reaction path and product analysis of sodium-water chemical reactions using laser diagnostics

Deguchi, Yoshihiro*; Muranaka, Ryota*; Kamimoto, Takahiro*; Takagi, Taku*; Kikuchi, Shin; Kurihara, Akikazu

Proceedings of 3rd International Workshop on Heat Transfer Advances for Energy Conservation and Pollution Control (IWHT 2015) (CD-ROM), 6 Pages, 2015/10

The purpose of this study aims to clarify the gas phase sodium-water reaction path and reaction products quantitatively. The counter-flow diffusion experiment device was employed to analyze the reaction path and reaction products using laser diagnostics. The main product of sodium-water reaction was determined to be NaOH and the sodium oxide was not notably measured compared with NaOH.

Journal Articles

Numerical investigation of self-wastage phenomena in steam generator of sodium-cooled fast reactor

Jang, S.*; Takata, Takashi; Yamaguchi, Akira*; Uchibori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.4275 - 4288, 2015/08

Numerical analysis of the self-wastage phenomenon was carried out using a multi-dimensional computational code called SERAPHIM. Several steps of numerical analysis were constructed to reproduce transient self-wastage phenomenon caused by Sodium Water Reaction (SWR). Numerical analysis of multiphase flow with chemical reaction near the initial crack is firstly performed. The wastage amount is evaluated based on hypothetical Arrhenius equation by using the temperature and molar concentration of sodium hydroxide. New analytical grid is created by exchanging the solid cells to fluid cells in the reaction based on the wastage amount evaluation. These series of procedure is repeated. The width and the shape of the enlarged crack showed good agreement with the experimental results.

Journal Articles

Numerical quantification of self-wastage phenomena in sodium-cooled fast reactor

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.

JAEA Reports

Development of experimental method for self-wastage behavior in sodium-water reaction; Development of test rig (SWAT-2R) and study for experimental procedure

Abe, Yuta; Shimoyama, Kazuhito; Kurihara, Akikazu

JAEA-Technology 2014-026, 40 Pages, 2014/07

JAEA-Technology-2014-026.pdf:33.12MB

In case of water leak from a penetrated crack on a tube of steam generator in the sodium cooled fast reactor (SFR), self-wastage, that increases the size of leak, may take place by corrosion related to chemical reaction between sodium and water. For the safety evaluation of the accident, JAEA has been developing the analytical method of self-wastage using the multi-dimensional sodium-water reaction code. This report describes the development of new experimental rig (SWAT-2R). SEAT-2R enables to examine corrosion effecting factors that were ambiguous in the previous studies. The report includes description of development of micro-leak test piece, examination of experimental procedure. The results will provide fundamental data for validation of the self-wastage analytical method.

Journal Articles

Heat transfer characteristics of sodium-water reaction jet around a tube in steam generator of sodium-cooled fast reactor

Kurihara, Akikazu; Umeda, Ryota; Shimoyama, Kazuhito; Abe, Yuta; Kikuchi, Shin; Ohshima, Hiroyuki

Nippon 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.

Journal Articles

Development of numerical evaluation methods for multi-physics phenomena under tube failure accident in steam generator of sodium-cooled fast reactor

Uchibori, Akihiro; Kikuchi, Shin; Kurihara, Akikazu; Hamada, Hirotsugu; Ohshima, Hiroyuki

Nippon Kikai Gakkai Rombunshu, B, 79(808), p.2635 - 2639, 2013/12

Multi-physics analysis system for a heat transfer tube failure event in a steam generator of sodium-cooled fast reactors has been developed. In this study, applicability of the newly constructed numerical models in the analysis system was investigated. The droplet entrainment / transport model which was incorporated into the SERAPHIM code was verified through the analysis of the related experiment. The experimental data about the pressure variation when the droplet entrainment occurs was reproduced by our model successfully. The TACT code is integrated by the numerical models of fluid-structure thermal coupling, stress evaluation and failure judgment of the structure. The fluid-structure thermal coupling model could predict the temperature distribution formed by the flow around the circular cylinder. About the failure judgment model, the predicted time of failure occurrence showed good agreement with the results of the tube rupture simulation experiment.

Journal Articles

Reaction path analysis to sodium-water chemical reaction field using laser diagnostics

Tamura, Kenta*; Deguchi, Yoshihiro*; Muranaka, Ryota*; Kusano, koji*; Takata, Takashi*; Kikuchi, Shin; Kurihara, Akikazu

Proceedings of 24th International Symposium on Transport Phenomena (ISTP-24) (USB Flash Drive), 5 Pages, 2013/11

The purpose of this study aims to clarify the gas phase sodium-water reaction path and reaction products. The counter-flow diffusion experiment device is in the form of introducing the argon-based water vapor from the top of depressurized reaction chamber to the liquid sodium pool. Na, Na$$_{2}$$, H$$_{2}$$O, and reaction products in the counter-flow sodium-water reaction field were measured using laser diagnostics. The temperature controlled device was also improved to reduce the condensation of Na in the reaction zone for the better measurement performance. The main product in the sodium-water reaction was determined to be NaOH from the experimental results and its reaction path was discussed using Na-H$$_{2}$$O elementary reaction analysis.

152 (Records 1-20 displayed on this page)