Pin cladding materials

Austenitic steels Extensive R D work has been devoted to austenitic steels by all the countries developing fast reactors The status of development in the main national groups can be summarized as follows 

The post-irradiation results have also confirmed the excellent resistance of these materials to external sodium corrosion (maximum depth of 30 mm measured on a PHENIX pin irradiated to 13,6 at % over 826 EFPD (Equivalent Full Power Days) Recent development 

Russia. Conventional austenitic stainless steels have been optimized by addition of B, Ti and Mg. Extensive experience involving more than 100 000 fuel pins, has been gained on the 20% CW TchS-68 steel irradiated in BN-600 with oxide and MOX fuels at peak temperatures up to 700°C [Refs. 7.12,7.13]. More than 2500 fuel pins reached doses more than 80 dpa, with a maximum dose of 94 dpa. This steel has been used successfully as the standard cladding material since the beginning of 1991. Post-irradiation profilometry has 

Diametral deformation on vanous austenitic materials in the PHENIX reactor. 

Comparison of the tensile properties of the standard 15-15 Ti Phenix, 1.4970 standard and Si-modified 15-15 Ti irradiated to the same dose (115 dpa) Tests were performed at the irradiation temperature. 

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As known, after the 26-th run the BN-600 reactor core has been changed over to the third charge with a maximum fuel bum-up of 10 % h.a., the maximum dose being 75 dpa. The fuel subassembly duct material is 3ri-450 type ferritic - martensitic steel (13Cr-2Mo-Nb-P-B), the fiiel pin cladding material is the cold worked austenitic steel (16Cr-15Ni-2Mo-2Mn-Ti-P-B). 

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