Tsunemitsu Yoshitake

Burst properties of irradiated oxide dispersion strengthened ferritic steel claddings

The effects of fast neutron irradiation on the burst properties of oxide dispersion strengthened (ODS) ferritic steel claddings which were previously manufactured by warm working as the first trial cladding tube manufacturing in Japan Nuclear Cycle Development Institute, were investigated. The samples were irradiated in the experimental fast reactor JOYO using the core material irradiation rig (CMIR) at temperatures between 723 and 878 K to fast neutron fluences ranging from 2.1 to 4.2×1026 n/m2 (E>0.1 MeV). The burst tests were conducted on a total of four irradiation conditions. The result of burst tests showed that the burst strength of the irradiated claddings was higher than that of unirradiated at the test temperatures up to 873 K and that the diametrical strain just before rupture of irradiated specimens was almost similar to unirradiated one. It was suggested that there was no irradiation embrittlement under the irradiation conditions examined.

Effects of Fast Reactor Irradiation Conditions on Tensile and Transient Burst Properties of Ferritic/Martensitic Steel Claddings

The effects of fast neutron irradiation conditions have been investigated by focusing on the mechanical properties of 11Cr-0.5Mo-2W, Nb, V ferritic/martensitic (F/M) stainless steel (PNC-FMS) and 10.5Cr-1.5Mo, Nb, V F/M stainless steel (HT9M) claddings, especially tensile and transient burst properties. These two F/M claddings were irradiated at temperatures from 693 to 833K to 42.5 dpa (displacement per atom) in the experimental fast reactor JOYO using the PFB090 fuel test subassembly. Post-irradiation tensile and temperature-transient-to-burst tests were carried out for defueled cladding specimens. The results of mechanical tests for the PNC-FMS cladding showed that there was no significant degradation in tensile and transient burst strengths even after fast neutron irradiation. On the other hand, the strength of the HT9M cladding tended to shift to lower values than those of as-received specimens. The differences in tensile and transient burst strengths between the two claddings were attributed to martensite structural stability which was related to the stable solid solution elements.

Behavior of Irradiated Type 316 Stainless Steels under Low-Strain-Rate Tensile Conditions

The effects of lower strain rate on the tensile behavior of 12 % cold-worked type 316 stainless steels irradiated in the EBR-II reactor under low-dose-rate and moderate temperature conditions were investigated. Tensile tests were carried out at a strain rate of 1 × 10−7/s. Post-test fractography and microstructural characterization were also performed. Irradiation temperature and dose appeared to have the greatest effect on hardening and ductility loss, whereas dose rate appeared to have less apparent effects. In conjunction with earlier work performed on the same material at a strain rate of 1 × 10−3/s, there was no significant effect of strain rate on tensile behavior under the irradiation conditions examined. For fracture behavior, the material after irradiation exhibited typical ductile fracture during both high and low-strain-rate tests.

Swelling and Microstructural Evolution in 316 Stainless Steel Hexagonal Ducts Following Long-Term Irradiation in EBR-II

Swelling behavior and microstructural evolution of 12% cold-worked 316 SS hexagonal ducts following irradiation in the outer rows of EBR-II is described. Immersion density measurements and transmission electron microscopy (TEM) examination were performed on a total of seven irradiation conditions. The samples were irradiated to temperatures between 375 and 430 C to doses between 23 and 51 dpa and at dose-rates ranging from 1.3 x 10{sup -7} to 5.8 x 10{sup -7} dpa/s. Dose-rates and temperatures approach conditions experienced by a variety of components in pressurized water reactors (PWRs) and those which may be present in future advanced reactors designs. TEM analysis was employed to elucidate the effect of radiation on the dislocation, void and precipitate structures as a function of irradiation conditions. A moderate dose-rate effect was observed for samples which were irradiated at dose-rates differing by a factor of two. Lower dose-rate samples contained voids of larger diameter and typically swelled more in the bulk. The dislocation and precipitate structure was not visibly influenced by a dose-rate decrease.

Effects of Microstructural Evolution on Charpy Impact Properties of Modified Ferritic/Martensitic Steel after Neutron Irradiation

The effects of fast neutron irradiation have been investigated on Charpy impact properties and microstructural evolution of ferritic/martensitic (F/M) 11Cr-0.5Mo-2W, Nb, V stainless steel (PNC-FMS) which were developed by the Japan Atomic Energy Agency for core component applications in advanced fast reactors. One-half-size Charpy specimens were irradiated in the experimental fast reactor JOYO using the material irradiation rig to doses of 4.4 dpa at 773 K and 2.8 dpa at 923 K. Post irradiation Charpy impact tests and microstructural observations by transmission electron microscope were carried out. The results of the Charpy impact tests showed that there was no significant degradation in Charpy impact properties after neutron irradiation at 773 K to 4.4 dpa compared with the as-received specimens, but there was significant degradation in the properties after neutron irradiation at 923 K to 2.8 dpa. The Charpy impact property changes were attributed to matrix softening by recovery of martensite lath structures and to precipitate distributions.