Development of negative-ion based NBI system for JT-60

Kuriyama, Masaaki; Aoyagi, Tetsuo; ; *; Ito, Takao; Inoue, Takashi; Usami, H*; Usui, Katsutomi; ; Oshima, Katsumi*; et al.

Nihon Genshiryoku Gakkai-Shi, 38(11), p.912 - 922, 1996/00

https://doi.org/10.3327/jaesj.38.912

no abstracts in English

Wastage characteristics of high-chrome steel heat transfer tube; Intermediate leak wastage tests

Shimoyama, Kazuhito; Hamada, Hirotsugu; Tanabe, Hiromi; Usami, Masayuki

PNC TN9410 93-212, 134 Pages, 1993/09

PNC-TN9410-93-212.pdf:5.99MB

A one-through unit type steam generator (SG) having the Mod.9Cr-1MO Steel for its heat transfer tube is considered to be promising for the development of large FBR SGs. Wastage data of the tube material was already obtained for the micro-/small leak region as formerly reported. Therefore, intermediate leak wastage tests were conducted in the range from 10 g/s to around 200 g/s by using the SWAT-1 test facility and the test results are summarized as follows: (1)The wastage resistivity of the Mod.9Cr-1Mo steel is between that of 2.25Cr-1Mo steel and austenitic stainless steel; namely, the Mod.9Cr-1Mo steel has about half the of wastage rate of the 2.25Cr-1Mo steel. An experimental wastage formula in the intermediate leak region was derived from the test data. (2)Almost all of the wastage profile of target tubes was toroidal type and it became about half the cross section area of the 2.25Cr-1Mo steel. An experimental formula on initial leak diameters versus equivalent secondary failure diameters was derived in the intermediate leak region. These test results would be applied to failure propagation analysis code LBAP which is to be used for the design of a one-through unit type SG.

Integrity confirmation test for duplex-wall heat transfer tubes in case of its inner-wall leak

Hamada, Hirotsugu; *; Himeno, Yoshiaki; *

PNC TN9410 89-146, 85 Pages, 1989/08

PNC-TN9410-89-146.pdf:2.61MB

Steam wastage tests of the duplex-wall heat transfer tubes for the steam generators were conducted by placing its emphasis on the investigation of the possible occurrence of a subsequent failure on an outer-wall of the tube in case of its inner-wall leak. Based on the limitation from the test rig, the test tubes, each of which has an artificial crack on its outer-wall instead of on its inner-wall, were manufactured and were subjected to the test. In the test, a super-heated and pressurized steam as conceptual plant design or a nitrogen gas was fed to the tube and was impinged against the inner-wall through its crack for 24 hours which are conservative enough to evaluate test results. The test with a nitrogen gas was to obtain reference data. Before and after the test, equivalent hydraulic diameter of the crack was determined by measuring a pressure drop due to a flowing helium assuming that the crack can be regarded as an orifice. Then, changes in the equivalent diameter of the crack were determined. After the test, the tubes were subjected to the post-test metallurgical examination. Results thus obtained are as follows: (1)In all tubes, equivalent diameter of the cracks decreased after the test. Some cracks were even plugged by stream corrosion products. No enlargements of the crack, therefore, was found. (2)Post-test metallurgical examination showed no evidence of a steam wastage. Only steam corrosion products were found in the gap between the inner and outer walls. In conclusion, within the extent of the present test, failure possibility of the duplex-wall tube following a generation of an initial crack on an inner-wall is negligible.

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