Development of a fast reactor for minor actinides transmutation, 1; Overview and method development

Takeda, Toshikazu*; Usami, Shin; Fujimura, Koji*; Takakuwa, Masayuki*

Proceedings of 21st International Conference & Exhibition; Nuclear Fuel Cycle for a Low-Carbon Future (GLOBAL 2015) (USB Flash Drive), p.560 - 566, 2015/09

The Ministry of Education, Culture, Sports, Science and Technology in Japan has launched a national project entitled "technology development for the environmental burden reduction" in 2013. The present study is one of the studies adopted as the national project. The objective of the study is the efficient and safe transmutation and volume reduction of minor actinides with long-lived radioactivity and high decay heat contained in high level radioactive wastes by using sodium cooled fast reactors. We are developing MA transmutation core concepts which harmonize efficient MA transmutation with core safety. To accurately design the core concepts we have improved calculation methods for estimating the transmutation rate of individual MA nuclides, and estimating and reducing uncertainty of MA transmutation. The overview of the present project is first described. The method improvement is presented with numerical results for a minor-actinide transmutation fast reactor.

Investigation on the sodium leak accident of Monju; Sodium leak test simulating the Monju leak

Shimoyama, Kazuhito; Nishimura, Masahiro; Usami, Masayuki; Miyahara, Shinya; Miyake, Osamu; Tanabe, Hiromi

PNC TN9410 97-085, 163 Pages, 1996/11

PNC-TN9410-97-085.pdf:6.17MB

Sodium fire experiments were carried out two times using the Sodium Fire Test Rig (SOFT-1) in the Power Reactor and Nuclear Fuel Development Corp (PNC) as a part of works to research the cause of the accident in secondary main cooling system of Monju. The purposes of these experiments are to confirm the leak rate and leakage form of sodium from damaged thermometer, to confirm the damage to the piping insulating structure around the thermometer and to the flexible tube, and to compare the temperature history of the signal from the thermometer between the experiments and Monju. In the experiments 56(2)g/sec was obtained as the leak rate under the condition of ensuring the leakage pass in the simulated thermometer. This leak rate was corrected to 53g/sec to take account of manufacturing error of the theemometer between the experiment and Monju. In calculation of this leak rate, it is assumed that the annulus size of thermometer well tip is a nominal distance and pressure value to the leakage sodium is 1.65kg/cmG, which was the maximum one during the leakage of Monju. Concerning the leakage form, connection condition between the thermometer and flexiblc tube affected the dropping style of the leaking sodium especially in its initial behavior. For the connection condition of the thermometer and flexible tube at the beginning of the experiments, the first experiment was started removing the connection to simulate the post accident observation results of Monju, while the second one was started in connected condition. In the second experiment, the connection condition became to be equal with the initial state of the first experiment 17 seconds after the beginning of thc leak ; the cap nut which fixed the flexible tube to the elbow connector came off. Until the connection came off, the typical leakage form was the dispersion from the elbow connector as a droplet and the flow penetrating the covering of the flexible tube as a streamline, while after the ...

Wastage characteristics of welds of high-chrome steel heat transfer tube of steam generator; Micro-Leak and small leak sodium-mater reactions

Shimoyama, Kazuhito; ; Usami, Masayuki; Tanabe, Hiromi; Yoshida, Eiichi; *

PNC TN9410 91-288, 72 Pages, 1991/07

PNC-TN9410-91-288.pdf:2.3MB

For the development of a once-through unit type steam generator for an FBR, materials having a high mechanical strength at elevated temperatures and high resistivity against stress corrosion cracking (scc) are needed as its heat transfer tubes. High-chrome ferritic steels such as Mod.9Cr-1Mo, 9cr-2Mo, and 9Cr-1Mo-Nb-V steel are considered to meet these requirements. As previously reported, there was not large difference of resistivity among these materials and it was confirmed that high-chrome ferritic steels were more wastage -resistant than the 2.25Cr-1Mo steel which is a representative ferritic steel as tube materials. Since wastage data of weld portion at which the initial leak is apt to occur had been insufficient in order to totally evaluate the wastage resistivity of the high-chrome ferritic steels wastage tests in the micro-leak and small leak ranges were conducted for the tube-tube weld point of Mod.9Cr-1Mo steel. Major results are as follows : (1)In the micro-leak tests, there are not large differences in the self-wastage resistivity between the weld and the base metal. And the wastage resistivity is independant of the location of the leak starting point (a weld metal, a bond of weld, or a heat affected zone). (2)In the small leak tests, there are not large differences in the wastage resistivity between the weld and the base material. In the results, the empirical formula of wastage rate that derived from the base materials of the high-chrome ferritic steel can also be applied to the heat transfer tubes including the weld materials.

Development of a variable proton ratio ion source

Watanabe, Kazuhiro; Dairaku, Masayuki; ; Horiike, Hiroshi; Inoue, Takashi; Kitamura, Shigeru; Komata, Masao; Kurashima, Toru*; Mizuno, Makoto; Oga, Tokumichi; et al.

JAERI-M 88-022, 26 Pages, 1988/02

JAERI-M-88-022.pdf:0.79MB

no abstracts in English

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Uruwashi, Shinichi; ; Usami, Masayuki;

PATRAM'95, , 

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