High Temperature Creep-Fatigue Design and Service Experience
- PDF / 608,147 Bytes
- 10 Pages / 612 x 792 pts (letter) Page_size
- 104 Downloads / 162 Views
1125-R02-06
High Temperature Creep-Fatigue Design and Service Experience A-A. F. Tavassoli, B. Fournier, M. Sauzay Commissariat à l’Energie Atomique, DEN/DMN, 91191 Gif-sur-Yvette, France ABSTRACT Generation IV fission and future fusion reactors envisage development of more efficient high temperature concepts where materials performances under creep-fatigue hold the key to success. This paper presents extended experimental results obtained from creep, fatigue and creep-fatigue tests on the main structural materials retained for these concepts, namely: stainless steel type 316L(N), the conventional Modified 9Cr-1Mo martensitic steel and its low activation derivatives such as Eurofer steel, and their more advanced grades strengthened by oxide dispersion. It shows that the existing recommendations made in design codes adequately cover individual damage due to creep or fatigue but often fall short under combined creep-fatigue interaction. This is partly due to the difficulties of reproducing service conditions in laboratory. In this paper, results from tests performed on components removed from reactor, after long service, are used to refine code recommendations. Using the above combined assessment, it is concluded that there is good confidence in predicting creep-fatigue damage for austenitic stainless steels. For the martensitic steels the effects of cyclic softening and microstructure coarsening throughout the fatigue life need more consideration in creep-fatigue recommendation. In the long-term development of ferritic/martensitic oxide dispersion strengthened grades with stable microstructure and no cyclic softening, appears promising provided problems associated with their fabrication and embrittlement are resolved. INTRODUCTION International agreements on Generation IV fission and future fusion reactors all target development of more efficient high temperature reactors [1, 2]. The choice of concepts to investigate for Gen IV has now, at least in France, been narrowed down from the initial 6 to the sodium cooled fast reactor as the main option and a higher temperature gas cooled fast reactor (GFR) as alternative [3]. Likewise, the concepts to be tested in ITER for the future fusion demonstration (DEMO) reactor, at least in Europe, have been reduced to two blanket modules with an upper temperature limit of 550 °C, with possible extension to 650°C [4]. Both fission and fusion programmes consider in the much longer term (> 20 years), development of higher temperature reactors using more advanced materials such as tungsten alloys and SiCf/SiC ceramic composites [5]. These are not discussed here1. MATERIALS AND SERVICE CONDITIONS The sodium cooled fast reactor (SFR) operating conditions are similar to those used for the Superphenix and the European Fast Breeder Reactors (EFR), with the lower core structures 1
Fuel cladding and fuel subassemblies are not permanent structures and are not discussed here.
exposed to cooler sodium at around 400 °C and the upper core structures to higher temperatures at around 550°C [3, 6
Data Loading...
