The structures of Reactor Pressure Vessel Internals are subjected to an intense neutron flux. Under these operating conditions, the microstructure and the mechanical properties of the austenitic stainless steel components change. In addition, these components are subjected to stress of either manufacturing origin or generated under operation. Cases of baffle bolts cracking have occurred in CP0 Nuclear Power Plant units. The mechanism of degradation of these bolts is Irradiation-Assisted Stress Corrosion Cracking. In order to obtain a better understanding of this mechanism and its principal parameters of influence, a set of stress corrosion tests (mainly constant load tests) were launched within the framework of the EDF project "Lower Core Internals," using materials from a CHOOZ A baffle corner (SA 304). These tests aim to quantify the influence on IASCC of the applied stress, temperature and environment (primary water, higher lithium concentration, inert environment) for an irradiation dose close to 30 dpa. A curve showing time to failure as a function of the stress was determined. The shape of this curve is consistent with the few data that are available in the literature. A stress threshold of about 50 % of the yield strength value at the test temperature has been determined, below which cracking in that environment seems unlikely. After irradiation this material is sensitive to intergranular fracture in a primary environment, but also in an inert environment (argon) at 340°C. The tests also showed a negative effect of increased lithium concentration on the time to failure and on the stress threshold.
|Title of host publication||Fontevraud 7|
|Subtitle of host publication||Contribution of Materials Investigation to Improve the Safety and Performance of LWRs, Avignon, France, 26-30.9.2010|
|Number of pages||18|
|Publication status||Published - 2010|
|MoE publication type||Not Eligible|
- stainless steel