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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

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.

Oral presentation

Study on sodium-water reaction phenomena in steam generator of sodium-cooled fast reactor, 38; Development of long-term leak enlargement and propagation analysis code

Hamada, Hirotsugu; Uchibori, Akihiro; Ohshima, Hiroyuki

no journal, , 

For the purpose of a safety evaluation of heat transfer tube failure in the FBR steam generator, the long-term leak enlargement and propagation analysis code (LEAP-III) is under development. A model of overheating tube rupture was incorporated into LEAP-III and LEAP-III was applied to an analysis of SWAT-3 test to evaluate the applicability of the code.

Oral presentation

Rapid heating tube rupture simulation experiments in case of sodium-water reaction in steam generator of sodium-cooled fast reactor

Umeda, Ryota; Kurihara, Akikazu; Shimoyama, Kazuhito

no journal, , 

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 material strength standard which is one of the major influencing factor. Therefore, in present study, the authors carried out tube rupture experiments with rapidly-heating which were simulated the tube thermally-affected by sodium-water reaction jet, and evaluated quantitatively failure hoop stress and failure time. Then, the authors confirmed that existing stress strength standard was applicable to thin diameter and thick-walled single tube in case of sodium-water reaction exceeding 1300$$^{circ}$$C under practical steam generator operation conditions.

Oral presentation

Tube rupture simulation experiments on the sodium-water reaction in steam generator of sodium-cooled fast reactor

Umeda, Ryota; Kurihara, Akikazu; Shimoyama, Kazuhito

no journal, , 

no abstracts in English

Oral presentation

Rapid heating tube rupture simulation experiments in case of sodium-water reaction in steam generator of sodium-cooled fast reactor, 2

Kurihara, Akikazu; Umeda, Ryota; Shimoyama, Kazuhito

no journal, , 

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 material strength standard which is one of the major influencing factor. In this report, rapid heating tube rupture experiments were conducted on the double-walled tube (Mod.9Cr-1Mo steel) under the same experimental conditions as those in the single-walled tube experiments, and evaluated quantitatively failure hoop stress and failure time. The authors confirmed the validity of the existing stress strength standard under the conditions that the double-walled tubes were uniformly heated.

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