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Initiation Stress Threshold Irradiation Assisted Stress Corrosion Cracking Criterion Assessment for Core Internals in PWR Environment

[+] Author Affiliations
Benoit Tanguy, Anthony Stern, Philippe Bossis

CEA, Gif-sur-Yvette, France

Cédric Pokor

EDF les Renardières, Moret-sur-Loing, France

Paper No. PVP2011-58051, pp. 681-687; 7 pages
  • ASME 2011 Pressure Vessels and Piping Conference
  • Volume 6: Materials and Fabrication, Parts A and B
  • Baltimore, Maryland, USA, July 17–21, 2011
  • Conference Sponsors: Pressure Vessels and Piping Division
  • ISBN: 978-0-7918-4456-4
  • Copyright © 2011 by ASME


Irradiation assisted stress corrosion cracking (IASCC) is a problem of growing importance in pressurized water reactors (PWR). An understanding of the mechanism(s) of IASCC is required in order to provide guidance for the development of mitigation strategies. One of the principal reasons why the IASCC mechanism(s) has been so difficult to understand is the inseparability of the different IASCC potential contributors evolutions due to neutron irradiation. The potential contributors to IASCC in PWR primary water are: (i) radiation induced segregation (RIS) at grain boundaries, (ii) radiation induced microstructure (formation and growth of dislocations loops, voids, bubbles, phases), (iii) localized deformation under loading, (iv) irradiation creep and transmutations. While the development of some of the contributors (RIS, microstructure) with increasing doses are at least qualitatively well understood, the role of these changes on IASCC remains unclear. Parallel to fundamental understanding developments relative to IASCC, well controlled laboratory tests on neutron irradiated stainless steels are needed to assess the main mechanisms and also to establish an engineering criterion relative to the initiation of fracture due to IASCC. First part of this study describes the methodology carried out at CEA in order to provide more experimental data from constant load tests dedicated to the study of initiation of SCC on neutron irradiated stainless steel. A description of the autoclave recirculation loop dedicated to SCC tests on neutron irradiated materials is then given. This autoclave recirculation loop has been started on July 2010 with the first SCC test on an irradiated stainless steel (grade 316) performed at CEA. The main steps of the interrupted SCC tests are then described. Second part of this paper reports the partial results of the first test performed on a highly neutron irradiated material.

Copyright © 2011 by ASME



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