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Dai-8-Kai Jikken Rikigaku Kokusai Kaigi, 0 Pages, 1996/06
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Umeda, Hisao;
PNC TN9410 93-156, 147 Pages, 1993/06
(1)[0BJECT] The object of this test is to obtain structural failure data of a typical structural model representing FBR components subjected to thermal loadings in order to develop more rationalized structural strength evaluation methods. This report provides the experimental results. (2)[TEST MODEL] The Welded Vessel Model has noteworthy typical shapes and stress distributions as can be found in the structural design of Large Type Fast Breeder Reactor, and the modified austenitic stainless steel, SUS316FR, which is hopeful for LFBR is incorporated. (3)[THERMAL TRANSIENT TEST CONDITION] Thermal creep-fatigue test was conducted with the Thermal Transient Test Facility for Structures (TTS). The test model was subjected to cyclic thermal transient of 250 C - 600 C by sodium. The cycle time of one thermal transient was 3 hrs, in which 250C sodium flowed into the model for 1 hr and 600 C sodium for 2 hrs. The thermal transient was as severe as the by temperature change rate of 40 C/sec. (4)[TEST RESULTS] The Thermal Transient Test was completed at 1055 cycle due to crack observation on the upper Y-junction. (5)[TEMPERATURE TRANSIENT] The temperature data was obtained which was necessitated for the subsequest thermal stress analysis. The observed temperature transients were steeper than analyses. (6)[CRACK OBSERVASION] Observation method with use of "STRAW SCOPE" during the test was effective for identification of crack initiation.
Umeda, Hisao; Tanaka, Nobuyuki; Watashi, Katsumi; Kikuchi, Masayuki; Iwata, Koji
Nuclear Engineering and Design, 140, p.349 - 372, 1993/06
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; Umeda, Hisao; ; ;
PNC TN9410 92-116, 174 Pages, 1992/01
The design and fabrication of the Filletted vessel Thermal Bending Model which is to be tested at TTS are described in this report. The objective of this model is to obtain the thermal transient strength of biaxial stress condition, which appears at Y-piece structures or the portions in the vicinity of sodium surface, and fillet welding which are noteworthy portions from the structural integrity viewpoint of Fast Breeder Reactor (FBR) components. As the testing portion, this model has three kinds of skirt structure, a thick walled cylinder and four stabilizer plates which are attatched on inner shell by fillet welding. The objective of the skirt structures and the thick walled cylinder is to investigate the relations between stress ratio, hoop and axial stress, and thermal transient strength, and they are designed in order to have same maximum principal stress range and large variation in the ratio of hoop to axial stress. The model has a straw-bag-like body which constituted with filleted cylinders, a thick walled cylinder and upper/lower trisphere dished plate. The model is supported with support cone at lower trisphere dished plate. The model has inner shell which strive for sodium flow stability.
Umeda, Hisao; ; ;
PNC TN9410 91-253, 221 Pages, 1991/01
The objective of this test is to obtain structural failure data of a typical structural model represening FBR components subjected to thermal loadings, and thereby to develope structural strength evaluation methods. This report provides the test results and crack observasion on the Stress Mitigation Structure Model(2). Thermal creep-fatigue test was conducted with the Thermal transient Test Facility for Structures(TTS). The test was successfuly performed, and we observed creep-fatigue cracks at all expected portions. The flow straightner which is a tested portion for confirmation of function appeared to be safety. Cracks were found on the surfaces of perforated plate and also inside of the tube-to-perforated plate weldments. There was no corrosion althogh sodium adhered to the joining face. The Performance of a thermal insulator could not be confirmed because sodium flowed inside it. The ultrasonic examination before inspection was effective for fine cracks even though it was impossible to catch exact depth. The temperature data and creep-fatigue cracks, that can be utilized to develop the design methods of the FBR components, could be obtained.
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PNC TN9410 89-170, 194 Pages, 1989/06
This report describes the result of thermal creep-fatigue test conducted at the Thermal Transient Test Facility for Structures. The model, named 'Thermal stress mitigation model 1', is subjected to cyclic thermal transient of 250 C-620C by sodium. The cycle time of one thermal transient is 2hrs, on which 250 C sodium flows into the model in 40 min and-620C sodium on 80 min. The model has six portions to be tested until failure by creep-fatigue, namely two nozzles, slitted cilinder, skirt structure, and four sorts of weldment. The test was successfuly performed, and we observed expected creep-fatigue crackings at all portions of interest. The temperature data and failure data were obtained for subsequent thermal stress analysis and fracture mechanics analysis. Also ultrasonic examination during the test was effective for identification of crack initiation an the hair crack stage.
*; *; *; Kasahara, Naoto; *; ; Imazu, Akira
PNC TN9410 89-088, 187 Pages, 1989/06
Thermal transient strength tests of structure models are carried out to develop the design method of the fast breeder reactor components under thermal loadings. The fifth testing model for Thermal Transient, Test Facility for Structures (TTS), "Thermal stress mitigation structure model (2)", have been designed and fabricated. The purpose of this model is to get the thermal transient strength data for the typical shape of FBR components and to confirm the function of specific structures under thermal loading. This testing model is a vertical type cylindrical vessel supported by a skirt. It has seven testing portion for failure test, such as two types of nozzle, a Y-junction, two types of skirt and a plate to shell junction. And it has three testing portion for confirmation of function, such as a thermal insulator, a flow straightener and tube to perforated plate weldments. In designing the model, thermo-hydraulic analysis, heat transfer analysis, thermal Stress analysis were performed. Testing portions were evaluated using the design guide for TTS exclusive use. Material and welding method are basically comparable to the prototype reactor internals.
; *; *; Nakanishi, Seiji; *
9th International Conference on structural Mechanice in Reactor Technology (9th SMIRT), ,
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Tanaka, Nobuyuki; Watashi, Katsumi; Umeda, Hisao; Kikuchi, Masayuki; Iwata, Koji
Int Symp on Structral Mechanics in Reactor Technology, ,
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