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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.

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

Reactive wetting by liquid sodium on thin Au plating

Kawaguchi, Munemichi; Hamada, Hirotsugu

Journal of Nuclear Science and Technology, 51(2), p.201 - 207, 2014/02

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

The behavior of sodium wetting is investigated for the practical use of the under sodium viewer (USV), in which the modeling of the reactive and non-reactive wetting for the metallic-plated steels by liquid sodium are performed to simulate the behavior of sodium wetting. Simulation results of the non-reactive wetting showed a good agreement with the Tanner's law. For the simulation of reactive wetting, the model of fluid flow induced by the interfacial reaction was incorporated into that of the non-reactive wetting. The simulation results of the reactive wetting, such as the behavior of precursor liquid film and the central droplet, showed a good agreement with the sodium wetting experiments using thin Au plating at 250$$^{circ}$$C. In reactive wetting simulation, it is important that the gradient of reaction energy at the interface appeared on the new interface around the triple junction, and that the fluid flow was induced.

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

Multiphysics analysis system for tube failure accident in steam generator of sodium-cooled fast reactor

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

Proceedings of 21st International Conference on Nuclear Engineering (ICONE-21) (DVD-ROM), 6 Pages, 2013/07

Multiphysics analysis system was newly developed to evaluate possibility of failure propagation occurrence under tube failure accident in a steam generator of sodium cooled fast reactors. The system consists of the computer codes, SERAPHIM, TACT, RELAP5, which are based on the mechanistic numerical models. The SERAPHIM codes calculates multicomponent multiphase flow involving sodium-water chemical reaction. In this study, a numerical model for chemical reaction about production of a sodium monoxide and its transport process were constructed. We also developed the numerical models of the TACT code for evaluation of shell-side flow around an adjacent tube, heat transfer from the fluid to the tube and occurrence of tube failure. In our analysis system, thermal hydraulic behavior of water inside the tube is evaluated by the RELAP5 code. The original heat transfer correlations were corrected for the rapidly heated tube in the present work.

JAEA Reports

Numerical methods on flow instabilities in steam generator

Yoshikawa, Ryuji; Hamada, Hirotsugu; Ohshima, Hiroyuki; Yanagisawa, Hideki*

JAEA-Research 2008-058, 29 Pages, 2008/06

JAEA-Research-2008-058.pdf:1.31MB

In Japan Atomic Energy Agency, investigations on heat transfer characteristics of steam generator are being performed for the development of Sodium-cooled Fast Breeder Reactor. In this report, the numerical methods were studied for two-phase flow instability analysis in steam generator. For numerical simulation purpose, the flow instability analysis code was developed with homogeneous equilibrium model on single heat transfer tube. The special algorithm to calculate inlet flow rate with inlet pressure, outlet pressure and heat flux as boundary conditions for the density-wave instability analysis has been established. The flow instability in single tube was successfully simulated with homogeneous equilibrium model. Then the drift-flux model including the effects of subcooled boiling and two phase slip was adopted to improve the accuracy. The capability of drift-flux model for simulating density-wave instability in single tube was confirmed.

JAEA Reports

Construction of interfacial area concentration model for sodium-water reaction

Yoshikawa, Ryuji; Ohshima, Hiroyuki; Hamada, Hirotsugu; Kurihara, Akikazu; Uchibori, Akihiro

JAEA-Research 2008-055, 24 Pages, 2008/06

JAEA-Research-2008-055.pdf:3.19MB

In Japan Atomic Energy Agency, thermal hydraulic studies on sodium-water reaction are being performed with the multi-component and multi-phase code SERAPHIM. The interfacial area concentration of sodium droplets in the steam is important for the accurate analysis of sodium-water reaction. In this report, the theoretical analysis and numerical models for gas jets were reviewed to understand the mixing process of sodium and water. As for theoretical analysis, existing critical flow rate, depressurization and entrainment analysis for jet flows were summarized. The applicability of critical flow rate equations for subcooled water at 17MPa were confirmed after investigating its effect of compressibility. Based on the available knowledge on entrained droplet sizes in gas jets, a transport equation of sodium droplet interfacial area concentration was constructed for multiphase flow simulation.

Journal Articles

Development of thermal hydraulic computer code for steam-water flow in steam generator of fast breeder reactor

Yoshikawa, Ryuji; Hamada, Hirotsugu; Ohshima, Hiroyuki; Yanagisawa, Hideki*

Dai-13-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.495 - 496, 2008/06

In Japan Atomic Energy Agency, investigations on heat transfer characteristics of steam generator are being performed for the development of commercialized sodium-cooled fast breeder reactor. In this study, the computer code for flow instability analysis in single heat transfer tube was developed with drift-flux model which included the effects of subcooled boiling and two phase slip. The special algorithm to calculate inlet flow rate with inlet pressure, outlet pressure and heat flux as boundary conditions for the density-wave instability analysis has been established. The subcooled model was verified by calculating the void fraction distribution of steady heat transfer flow. The capability of drift flux model for simulating density-wave instability in single tube was confirmed.

Journal Articles

Development of blow down and sodium-water reaction jet analysis codes; Validation by sodium-water reaction tests (SWAT-1R)

Seino, Hiroshi; Jitsu, Koji*; Kurihara, Akikazu; Ono, Isao*; Hamada, Hirotsugu

Proceedings of 13th International Conference on Nuclear Engineering (ICONE-13), 0 Pages, 2005/05

Blow down analysis code (LEAP-BLOW) and sodium-water reaction jet analysis code (LEAP-JET) have been developed to improve the evaluation accuracy on sodium-water reaction. The validation analyses by these codes were carried out using the data of SWAT-1R test. As the result, though there was a problem in the quantitative evaluation of LEAP-JET, it was possible to obtain the approximately appropriate results.

JAEA Reports

In-vessel source term analysis code TRACER version 2.3 user's manual

Toyohara, Daisuke*; Ohno, Shuji; Matsuki, Takuo*; Hamada, Hirotsugu; Miyahara, Shinya

JNC-TN9520 2004-004, 151 Pages, 2005/01

JNC-TN9520-2004-004.pdf:139.32MB

A computer code TRACER (Transport Phenomena of Radionuclides for Accident Consequence Evaluation of Reactor) version 2.3 has been developed to evaluate the species and the quantities of fission products (FPs) released into cover gas during fuel pin failure accident in an LMFBR. TRACER version 2.3 includes new and modified models shown below. a) Booth model, a new model for FPs release from fuel. b) Modified model for FPs transfer from fuel to bubble or sodium coolant. c) Modified model for bubbles dynamics in coolant. Computer models, input data and output data of TRACER Version 2.3 are described in this user's manual.

JAEA Reports

Equilibrium evaporation test of lead-bismuth eutectic and of tellurium in lead-bismuth

Ohno, Shuji; Nishimura, Masahiro; Hamada, Hirotsugu; Miyahara, Shinya; Sasa, Toshinobu*; Kurata, Yuji*

JNC-TN9400 2004-072, 52 Pages, 2005/01

JNC-TN9400-2004-072.pdf:2.88MB

A series of equilibrium evaporation experiment was performed to acquire the essential and the fundamental knowledge about the transfer behavior of lead-bismuth eutectic(LBE) and impurity tellurium in LBE from liquid to gas phase. The experiments were conducted using the transpiration method in which saturated vapor in an isothermal evaporation pot was transported by inert carrier gas and collected outside of the pot. The size of the used evaporation pot is 8cm inner diameter and 15cm length. The weight of the LBE pool in the pot is about 500g. The investigated temperature range was 450deg-C to 750deg-C. From this experiment and discussion using the data in literature, we have obtained several instructive and useful data on the LBE evaporation behavior such as saturated vapor pressure of LBE, vapor concentration of Pb, Bi and Bi2 in LBE saturated gas phase, and activity coefficient of Pb in the LBE. The LBE vapor pressure equation is represented as the sum of Pb, Bi and Bi2 vapor in the temperature range between 550deg-C and 750deg-C as logP[Pa]=10.2-10100/T[K]. The gas-liquid equilibrium partition coefficient of tellurium in LBE is in the range of 10 to 100, with no remarkable temperature dependency between 450deg-C and 750deg-C.

JAEA Reports

Lead-Bismuth Transfer Behavior Preliminary Test in Liquid Sodium; Effect of Test Temperature and Amount of Lead-Bismuth on Reaction Behavior

Saito, Junichi; Sagawa, Norihiko; Ohno, Shuji; Hamada, Hirotsugu; Miyahara, Shinya

JNC-TN9400 2004-059, 133 Pages, 2004/09

JNC-TN9400-2004-059.pdf:6.05MB

The simplified secondary sodium cooling system, in which utilized lead-bismuth eutectic is utilized as an intermediate coolant has been selected as one of candidate system for the "Feasibility Studies on Commercialized Fast Reactor System (Phase I)". The purpose of this study for the "FS (PhaseII)" is to understand transfer behavior of lead-bismuth in liquid sodium by experiment. The experiments which trickles liquid lead-bismuth into liquid sodium are carried out of under various test temperature and amount of lead-bismuth. The effects of test temperature and amount of lead-bismuth on reaction behavior of sodium and lead-bismuth are clarified from the experimental results. The obtained results from experiments are as follows. (1) The experiment under lower test temperature takes longer time for reaction between sodium and lead-bismuth than that under higher test temperature. It means that test temperature affects the reaction behavior between sodium and lead-bismuth. (2) The amount of dropping lead-bismuth affects an amount and kind of reaction products which are formed by reaction between sodium and lead-bismuth. (3) Reaction heat obtained from the experiments is similar to estimated reaction heat using formation enthalpy of BiNa3 which is one of dominant reaction products.

JAEA Reports

The Development and Application of Overheating Failure Model of FBR Steam Generator Tubes (4)

Miyake, Osamu; Hamada, Hirotsugu; Tanabe, Hiromi; Wada, Yusaku; Miyakawa, Akira; Okabe, Ayao; Nakai, Ryodai; Hiroi, Hiroshi

JNC-TN2400 2003-003, 225 Pages, 2004/02

JNC-TN2400-2003-003.pdf:40.45MB

The model has been developed for the assessment of the overheating tube failure in an event of sodium-water reaction accident of fast breeder reactor's steam generators (SGs). The model has been applied to the Monju SG studies.

Journal Articles

Analysis of Overheating Rupture in Heat-Transfer Tubes Causing Corrosive High-Temperature Reaction

HAMADA, Hirotsugu; TANABE, Hiromi;

Journal of Nuclear Science and Technology, 41(6), 665 Pages, 2004/00

 Times Cited Count:1 Percentile:88.58(Nuclear Science & Technology)

Sodium-water reaction tests simulating intermediate water leaks into sodium are analyzed by the overheating rupture model of a heat-transfer tube exposed to the corrosive and high-temperature reaction jet. The comparison of the model with test data leads to the following conclusions: The failure behaviors of gas-pressurized tubes are classified into two types of the creep failure and the ductile failure accompanied by creep, depending on the test conditions. Thin-wall tubes fail within tens of seconds due to the ductility while thick-wall tubes fail in about 1 minute due to the creep; in the latter case, the thinning due to wastage proved to be dominant in the overheating rupture phenomena. In the creep failure and the ductile failure accompanied by creep, the times to failure in the analysis are respectively estimated 20-50 % and 35-50 % shorter than those in the experiments; namely, the analysis has a certain degree of conservatism. In the creep failure, if the time coefficient

Journal Articles

Study of thermal influence on tubes due to sodium-water reactions in LMFBR steam generator

; HAMADA, Hirotsugu; Kurihara, Akikazu; Nishimura, Masahiro

Proceedings of 12th International Conference on Nuclear Engineering (ICONE-12) (CD-ROM), 0 Pages, 2004/00

A study of thermal influence on heat-transfer tubes in sodium-water reactions is carried out to precisely evaluate the tube rupture due to overheating in the water leak accident of an LMFBR steam generator (SG). By assuming the sodium-water reaction jet to be a two-phase flow that consists of sodium and hydrogen, the heat-transfer characteristics are examined and a simple model of effective heat-transfer coefficient (HTC) is proposed for the safety evaluation in the SG. Comparison of the model with experimental data leads to the following conclusions: An upper limit exists in the HTC between reaction jet to tube wall, and it is equivalent in approximation to the HTC of single-phase sodium flow. The HTC can be written in simpler form as functions of the HTC of single-phase sodium flow, void fraction and temperatures of sodium, hydrogen and tube wall. Hydrogen provides negligible heating effect, so that the apparent HTC would decrease with increasing of the hydrogen temperature that

JAEA Reports

Lead-Bismuth Transfer Behavior Preliminary Test in Liquid Sodium

Saito, Junichi; Takai, Toshihide; Sagawa, Norihiko; Ohno, Shuji; Hamada, Hirotsugu; Miyahara, Shinya

JNC-TN9400 2003-057, 87 Pages, 2003/06

JNC-TN9400-2003-057.PDF:24.73MB

The simplified secondary sodium cooling system utilized lead-bismuth eutectic as an intermediate coolant has been selected one of candidate system for the "Feasibility Studies on Commercialized Fast Reactor System (Phase I)". The purpose of this study for the "FS (Phase II)" is to understand transfer behavior of lead-bismuth in liquid sodium by experiment. From the experimental results the fundamental data are obtained for the feasibility evaluation of the simplified secondary sodium cooling system. Twice experiments which trickles liquid lead-bismuth into liquid sodium are carried out at 400 degress centigrade are obtained. (1) From the ICP analyses of L1-1 and L1-2 test, the lead concentration of sodium is higher than the bismuth concentration. This shows that the amount of dissolution of lead into liquid sodium was larger than that of bismuth. This result agrees with data of the previous solubility data in Pb-Na and Bi-Na binary system in sodium. The solid black particles observed in sodium contain a large amount of bismuth. (2) Temperature of liquid sodium rises when the drops of liquid lead-bismuth are added into liquid sodium. The total heating value calculated using temperature rises observed at several parts in equipment is 137 kJ/mol-LBE on L1-2 test. This heat of reaction is promising for leak detection of lead-bismuth into sodium. (3) Many black solid products are observed in sodium taken from L1-1 and L1-2 test apparatus. The reaction products taken from upper location in a sampling finger are very fine and the size was 5 ~ 10$$times$$10$$^{-6m.}$$Those from lower location increase in size and the size was 50 ~100$$times$$10-6m.

JAEA Reports

The Report of inspection and repair technology of sodium cooled reactors

Kisohara, Naoyuki; Uchita, Masato; Konomura, Mamoru; Kasai, Shigeo; Soman, Yoshindo; Shimakawa, Yoshio; Hori, Toru; Chikazawa, Yoshitaka; Miyahara, Shinya; Hamada, Hirotsugu; et al.

JNC-TN9400 2003-002, 109 Pages, 2002/12

JNC-TN9400-2003-002.pdf:8.12MB

Sodium is the most promising candidate of an FBR coolant because of its excellent properties such as high thermal conductivity. Whereas, sodium reacts with water/air and its opaqueness makes it difficult to inspect sodium components. These weaknesses of sodium affect not only plant safety but also plant availability (economy). To overcome these sodium weak points, the appropriate countermeasure must be adopted to commercialized FBR plants. This report describes the working group activities for sodium/water reaction of steam generators (SG), in-service inspection for sodium components and sodium leak due to sodium components boundary failure. The prospect of each countermeasure is discussed in the viewpoint of the commercialized FBR plants. (1)Sodium/water reaction. The principle of the countermeasure for sodium/water reaction accidents was organized in the viewpoint of economy (the investment of SG and the plant availability). The countermeasures to restrain failure propagation were investigated for a large-sized SG. Preliminary analysis revealed the possibility of minimizing tubes failure propagation by improving the leak detection system and the blow down system. Detailed failure propagation analysis will be required and the early water leak detection system and rapid blow down system must be evaluated to realize its performance. (2)In-service inspection (ISI&R). The viewpoint of the commercialized plant's ISI&R was organized by comparing with the prototype reactor's ISI&R method. We also investigated short-term ISI&R methods without sodium draining to prevent the degrading of the plant availability, however, it is difficult to realize them with the present technology. Hereafter, the ISI&R of the commercialized plants must be defined by considering its characteristics. (3)Sodium leak from the components. This report organized the basic countermeasure policy for primary and secondary sodium leak accidents. Double-wall structure of sodium piping was ...

JAEA Reports

Evaluation of Tube Rupture Simulation Test (TRUST-1) for FBR steam generators

Hayashida. Y; Hamada, Hirotsugu

PNC-TN9410 97-002, 38 Pages, 1996/06

PNC-TN9410-97-002.pdf:0.87MB

The intermediate water leak in an FBR Steam Generator (SG) causes a high temperature and corrosive sodium-water reaction jet. In such cases, it is necessary to evaluate the wastage and overheating rupture behavior of heat transfer tubes. Especially, in the large SG that aims at high temperature of sodium and high temperature/pressure of water, the establishment of the rational evaluation method is important. In this paper, as a basic experiment to make clear the phenomenon of overheating rupture, tests and analysis of Tube Rupture Simulation Test-1 (TRUST-1) were conducted. TRUST-1 simulates the overheating rupture of the tube made of Mod.9Cr-1Mo steel by nitrogen gaspressurization and quick induction heating. The result of TRUST-1 are as follows: (1)The breaking strength predicted by the internal pressure is larger than the tensile strength of the tube material. (2)The margin of the breaking strength from the tensile strength of the tube material has a tendency of decreasing with the heating rate, especially in the lower temperature region. (3)Using an theoretical formula that is deduced from the steady creep model and appropriate experimental coefficients that are determined by the test data, the breaking strength can be reasonably evaluated.

Oral presentation

Study of water leak in FBR steam generator aiming at higher sodium temperature, First report

Hamada, Hirotsugu; Miyake, Osamu; Miyahara, Shinya

no journal, , 

The intermediate water leak test was conducted with simulating the future steam generator conditions of unit type and higher sodium temperature, and the temperature distribution of sodium-water reaction jet and its effect on the neighboring tubes were studied. The results of the detailed data analysis and the verification of analysis code will be described in the following reports.

Oral presentation

Wastage characteristics caused by sodium-water reaction jet

Hamada, Hirotsugu; Ohshima, Hiroyuki

no journal, , 

The wastage was modeled and the main parameter was derived with considering a sodium-water reaction jet that received the influence of water injection condition causes erosion to heat transfer tubes. It was applied to the data of small and intermediate water leak tests, and the qualitative behavior of wastage was reasonably explained.

35 (Records 1-20 displayed on this page)