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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.

JAEA Reports

Manufacture of ultrasonic thickness measurement apparatus

Oba, Toshihiro; Yanagihara, Takao; Kato, Chiaki; Hamada, Shozo

JAERI-Tech 2001-059, 36 Pages, 2001/09

JAERI-Tech-2001-059.pdf:7.8MB

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.

JAEA Reports

Simple evaluations of fluid-induced vibrations for steam generator tube arrays in advanced marine reactors (MRX, DRX)

Saito, Kazuo*; Ishida, Toshihisa

JAERI-Tech 2001-039, 25 Pages, 2001/06

JAERI-Tech-2001-039.pdf:0.94MB

no abstracts in English

JAEA Reports

The 3rd technological meeting of Tokai reprocessing plant

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$$^{th}$$, 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.

JAEA Reports

Improvement effect on corrosion under heat flux on nitric acid solutions of anti-IGC stainless steel and high Cr-W-Sr Ni base alloy

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

JAEA Reports

Ultra-High temperature strength properties on Mod.9Cr-1Mo steel

; 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$$^{circ}$$C to 1300$$^{circ}$$C. 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 1300$$^{circ}$$C. 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.

JAEA Reports

A feasibility study of the particle interaction method for the flow regimes with the chemical reaction; (Report under the contract between JNC and Toshiba Corporation)

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.

JAEA Reports

Evaluation of steam generator U-tube integrity during PWR station blackout with secondary system depressurization

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

JAEA Reports

Development on in-service inspection sytem for heat transfer tubes in the primary pressurized water cooler of the HTTR

Shinozaki, Masayuki; *; Furusawa, Takayuki

JAERI-Tech 99-064, 46 Pages, 1999/08

JAERI-Tech-99-064.pdf:3.55MB

no abstracts in English

JAEA Reports

The development and application of overheating failure model of FBR steam generator tubes

; *; *; *; Hiroi, Hiroshi*

PNC-TN9410 98-029, 122 Pages, 1998/05

PNC-TN9410-98-029.pdf:14.03MB

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 - 1200$$^{circ}$$C) 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.

JAEA Reports

None

*; *; ; *; *; Ito, Kenji

PNC-TJ2164 97-004, 38 Pages, 1997/10

PNC-TJ2164-97-004.pdf:3.34MB

JAEA Reports

Development of eddy-current in-service inspection system for FBR steam generator tubes; Establishment of the set parameters for off line data analysis

; ;

PNC-TN9410 97-087, 142 Pages, 1997/07

PNC-TN9410-97-087.pdf:5.29MB

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.

Journal Articles

Seismic test of a heat exchanger with a helically coiled tube bundle

Futakawa, Masatoshi; *; Takada, Shoji;

Nuclear Technology, 118(1), p.83 - 88, 1997/04

 Times Cited Count:1 Percentile:84.31(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Evaluation on materials performance of Hastelloy alloy XR for HTTR uses, 6; Tensile and creep properties of heat exchanger tube base materials and its welded-joints

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

Journal Articles

Development of an ECT probe having exciting coils in the shape of parallelograms

; *; Ara, Katsuyuki

Electromagnetic Nondestructive Evalution, 0, p.223 - 230, 1997/00

no abstracts in English

JAEA Reports

None

*; *; *; *; *; *

PNC-TJ9164 96-023, 1167 Pages, 1996/07

PNC-TJ9164-96-023.pdf:23.37MB

None

JAEA Reports

Overheating failure analysis of steam generator tubes II ; Overheating failure analysis of U.K.PFR superheater

;

PNC-TN9410 96-027, 41 Pages, 1995/12

PNC-TN9410-96-027.pdf:1.02MB

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.

JAEA Reports

0verheating failure analysis of steatm generator tubes; Validation analysis of explosive torch overheating test

PNC-TN9410 95-262, 35 Pages, 1995/09

PNC-TN9410-95-262.pdf:0.83MB

Neighboring tubes in an FBR Steam Generator (SG) would suffer from overheating if a sodium-water reaction jet were formed due to water leakage in the SG. On the safety aspect of the SGs, it is important to confirm that the neighboring tubes would not fail under such an overheating condition. An analytical model using the structural integrity analysis code, FINAS, has been prepared to evaluate the overheating failure and here an explosive torch overheating test was analyzed to validate the FINAS model. These experiments and analysis indicate that the overheating failure is closely associated with heat transfer coefficients (HTCs) of outer and inner tube wall and that the FINAS model conservatively predicts the overheating failure within acceptable accuracy. For making progress in further tests like an explosive torch test and its code validation, it would be required that sodium-water reaction experiments should be performed to provide the data on the HTCs, high pressurized and superheated steam should be supplied in the explosive torch test, and that a multidimensional analytical model should be developed to closely predict the temperature distribution in the axial(z-) and circumferential($$theta$$-) directions on the tube wall.

Journal Articles

Structural integrity test for heat transfer tube of intermediate heat exchanger

Kaji, Yoshiyuki; Ioka, Ikuo;

The 3rd JSME/ASME Joint Int. Conf. on Nuclear Engineering (ICONE),Vol. 1, 0, p.363 - 368, 1995/00

no abstracts in English

JAEA Reports

Ultra-high temperature tensile properties on Mod.9Cr-1M0, 2.25Cr-1Mo and SUS321 steel(I)

; Yoshida, Eiichi;

PNC-TN9410 94-262, 120 Pages, 1994/09

PNC-TN9410-94-262.pdf:6.07MB

This study clarified the tensie properties of Mod.9Cr-1Mo, 2.25Cr-1Mo and SUS321 steels at ultra-high temperature up to 1,200$$^{circ}$$C which will be used in analysys and evaluation of the tube burst in steam generators of fast breeder reaetors. (1)Tensile strength of Mod.9Cr-1Mo, 2.25Cr-1Mo and SUS321 steels at 1,200$$^{circ}$$C were 2.5, 2,and 2.5kg/mm$$^{2}$$, respectively. (2)The difference for tensile strength and 0.2% yeild strength between specimen heat rate and heat holding time could not be found in the present. (3)The temperatures of the tube burst at the maximum internal pressure of 150kg/cm$$^{2}$$ corresponding to the practical use condition were expected to be approximately 960$$^{circ}$$C for Mod.9Cr-1M0, 860$$^{circ}$$C for 2.25Cr-1Mo and 1040$$^{circ}$$C for SUS321, respectively. These tests result will be reflected to evaluation of tube burst by sodium water reaction.

52 (Records 1-20 displayed on this page)