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Proceedings of 4th International Conference on Nuclear Engineering (ICONE-4), P. 561, 1996/00
INMM Annual Meeing, 35, 918 Pages, 1994/07
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PNC TN243 84-02, 77 Pages, 1984/09
Nuclear generated electric power in Japan attained to 1.05 x 1011KWH, about 22 % ofotal generated electric power, in 1983 fiscal year, because of LWR's satisfactory opation. It is expected thah LWR is continously the mainstream of nucelar power plantfor some time. 'Long-term prediction of the energy supply and demand' was revised ilast year by MITI. All of the energy supply and demand were decreased in comparisonith that of last prediction, reflecting a change of Japanese energy demand after theil shock. The total unclear power generation capacity also was decreased from about x 107 kW to about 6.2 x 107 kW in 2000. However, Fast Breeder Reactor (FBR) develoent has been one of most important elements of Japanese nuclear energy policy. Name, to establish nuclear fuel cycle utilizing Pu obtained by reprocessing is sequired r a energy security of Japan as a nation having no resource. Experimental FBR 'Joyohas been operating satisfactorily and prototype FBR 'Monju' has been preparin
*; *; Atsumo, Hideo
PNC TN241 81-23, 232 Pages, 1981/08
As a part of PNC-USDOE corporation in the field of fast breeder reactor, thermal transient test of 24-inch elbow, made of type 304 stainless steel, was performed at ETEC (Energy Technology Engineering Center). Test results were analysed and evaluated at PNC. Following results were drawn from the test and the relevant stress analysis: (1)Thermal transient test was performed under constant in-plane bending moment and cyclic rapid cooling by nitrogen gas. When the mechanical load was increased step by step, shakedown or ratchetting behavior was observed under or above a certain level of load. For typical points at inner and outer surface of elbow, strain was measured by using high-temperature strain gage. (2)Thermal ratchetting analysis was performed by using FINAS. Coincidence between analysis and test results was obtained for qualitative shake down or ratchetting behaviour. As to increment of displacement or strain per cycle, analytical results were fairly larger than test results. (3)In the case of this test, in which creep effect is negligible, ratchetting load calculated by current high-temperature structural design method (ASME Code Case N-47 and PNC's structural design guide) gives sufficiently concervative value. (4)Thermal ratchetting mechanism of elbow under in-plane bending moment was identified as follows by test and analysis : bending deformation in circumferential direction at elbow side; extensional deformation in longitudinal direction at outer point from the elbow side by 20-degree; compressive deformation in longitudinal direction at inner point from the elbow side by 30-degree. (5)Ratchetting condition can be estimated fairly well and conservatively by taking following parameters in Bree's diagram : x = [(membrane stress) + (bending stress)/(section factor)]/(yield stress), and y = (thermal stress using equivalent linear temperature difference)/(yield stress). (6)Comparing this test with another test for smaller elbow at O-arai Engineering ...
*; Iwata, Koji; *; *
PNC TN241 79-05, 354 Pages, 1979/01
For the structural design of FBR reactor components under high temperature, it is becoming more important to evaluate the creep rupture damage in detail. This report summarizes typical stress relaxation curves and creep rupture damage of a cylindrical model for austenitic stainless steel SUS304. Computer program TEPC, special purpose program to evaluate one dimensional stress behavior, is applied for this stress analysis.
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PNC TN241 78-04, 637 Pages, 1978/02
This report includes three reports on development of simplified inelastic analysis methods for prototype fast breeder reactor MONJU. The purpose of this work is to develop more practical and reasonable analysis method of high-temperature design of FBR components than simple but too restrictive elastic analysis method or detailed but too expensive and complicated inelastic analysis method. Three theme of the first priority were chosen and entrusted to the following companies. The resulted proposed simplified methods will be used for preliminary evaluation of structural components under the specified restricted load conditions. No.1 Development of Simplified Evaluation Method of Thermal Ratcheting for Piping Elbows (Hitachi, Ltd.) No. 2 Development of a Simplified Methods for Inelastic Analysis of Piping Systems for High Temperature Service (Ishikawajima-Harima Heavy Industries, Ltd.) No.3 Simplified Inelastic Analysis Method of Perforated Plates (Mitsubishi Heavy Industries, Ltd.)
