Measurements of physical properties of FBR (Fast Breeder Reactor) structural materials; Part 1 Measurements of physical properties of the rolled steels

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PNC TN9410 90-094, 80 Pages, 1990/06

PNC-TN9410-90-094.pdf:1.48MB

It is needed for FBR design to know physical properties of the structural materials that will be used for FBR-reactor vessels, -steam generators, and -pipes. In this report, six types of steels that are noted as the structural materials for next steps of FBR are taken up. The six types of steels are as follows ; [SUS 304] [SUS 316 (conventional)] [FBR grade SUS316] [SUS321] [2.25Cr-1Mo steel] [Mod.9Cr-1Mo steel (ASTM A387-91).] Physical properties of the steels are measured on rolled plates made by mills. The measured physical properties are as follows ; [Specific gravity] [Specific heat] [Thermal conductivity] [Thermal expansion] [Young's modulus] [Poisson's ratio.] And also the same kinds of physical properties of S.R. (Stess Relieving) heat treated 2.25Cr-1Mo and Mod.9Cr-1Mo steel plates are measured. Measurement of the physical properties of forgings, pipes, and weld metals of selected types of steels of them will be also carried out in 1990 and all summarized data that will contain the results of this report will be used to determin the criteria design values of the physical properties for the next steps of FBR. (Caution) This report only shows the measurement results of the physical properties of the rolled steel plates but doesn't give the design standards of the physical properties for the next steps of FBR.

Study of micro-defect self-wastage phenomena on LMFBR prototype steam generators' tube; Studies of micro-leak sodium-water reactions (2)

Kuroha, Mitsuo; *; *; *; *

PNC TN941 82-101, 185 Pages, 1982/04

PNC-TN941-82-101.pdf:8.79MB

A series of micro steam leak tests has been carried out to accumlate the micro defect self-wastage data on LMFBR prototype MONJU steam generators' tubes, whose materials are planned to be 2.25Cr-1Mo and 321 stainless steels. The SWAT-2 test loop was used for the first stage of tests, and the SWAT-4 test rigs manufactured exclusively for the tests have been used since the following tests at the O-arai engineering center in PNC. The six test pieces, each of which had a micro crack nozzle through the thickness, were tested under the various conditions; the sodium temperatures were 470C and 505C, the steam pressure was about 130 kg/cmg throughout the sixt tests, and the average steam leak rates were in the range of 210 to 410 g/sec before the self-enlargements. The Main results obtained from the six tests are as followed; (1)It was obserbed in the case of 2.25Cr-1Mo steel nozzle that though the average leak rate of about 2 10 g/sec was very small, the leakage continued without any self-plugs until the self-wastage was completed through the thickness and finally increased the leak rate, which can cause wastage damage on adjacent tubes. (2)Self-wastage rates were apparently dependent on materials, steam leak rates and sodium temperatures. However, the increased leak rates were not strongly affected by those test parameters, and whose maxium values were apt to be in the range from to 10 g/sec in spite of the different test parameters. (3)The metallurgical structure of the stainless steel nozzles enlarged in size at the self-wasted sections due to the sodium-steam reactions, however, the composition did not changed at all. The surfaces of the self-wasted sections were rugged near the sodium and steam sides, but comparatively even midway between the two sides.

Effect of Fast Neutron Irradiation on AISI316 Steel Tube

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Nihon Genshiryoku Gakkai-Shi, 14(6), p.283 - 289, 1972/00

https://doi.org/10.3327/jaesj.14.283

no abstracts in English