Low cycle fatigue in hot water

Routine practice in estimating behavior is described in Section III of the ASME code, which refers to design curves giving the number of allowable cycles as a function of alternating stress intensity. The design curves are plotted from rest results on test specimens. On examination of the conditions of these tests, it appears that most of them have been performed in environmental conditions which are different from those surrounding the materials of a nuclear reactor. This remark is consistent with the general policy of ASME code, of which the rules 'Do not cover deterioration which may occur in service as a result of radiation effects, corrosion, erosion, or instability of the Material'. In order to assess the deterioration of the material under the combined effect of cyclic load variations (fatigue) and the presence of pressurized hot water, it is therefore necessary to secure fatigue test results obtained in a medium which suitably represents the water in the circuit of a pressurized water reactor. This program is underway at Saclay, and the results already obtained have been confined to previous publications on this matter. (orig./GL)