148 NEUTRON-INDUCED MICROSTRUCTURAL EVOLUTION OF Fe-15Cr-16Ni ALLOYS AT ~ 400°C DURING NEUTRON IRRADIATION IN THE FFTF FAST REACTOR - T. Okita, T. Sato and N. Sekimura (University of Tokyo), F. A. Garner and L. R. Greenwood (Pacific Northwest National Laboratory)*, W. G. Wolfer (Lawrence Livermore National Laboratory), Y. Isobe (Nuclear Fuel Industries Ltd., Japan) OBJECTIVE The purpose of this effort is to determine the influence of dpa rate on void swelling of fcc alloys. SUMMARY An experiment conducted at ~400°C on simple model austenitic alloys (Fe-15Cr-16Ni and Fe-15Cr-16Ni-0.25Ti, both with and without 500 appm boron) irradiated in the FFTF fast reactor at seven different dpa rates clearly shows that lowering of the atomic displacement rate leads to a pronounced reduction in the transient regime of void swelling. While the steady state swelling rate (~1%/dpa) of these alloys is unaffected by changes in the dpa rate, the transient regime of swelling can vary from <1 to ~60 dpa when the dpa rate varies over more than two orders of magnitude. This range of dpa rates covers the full span of fusion, PWR and fast reactor rates. The origin of the flux sensitivity of swelling arises first in the evolution of the Frank dislocation loop population, its unfaulting, and the subsequent evolution of the dislocation network. There also appears to be some flux sensitivity to the void nucleation process. Most interestingly, the addition of titanium suppresses the void nucleation process somewhat, but does not alter the duration of the transient regime of swelling or its sensitivity to dpa rate. Side-by-side irradiation of boron-modified model alloys in this same experiment shows that higher helium generation rates homogenize the swelling somewhat, but do not significantly change its magnitude or flux sensitivity. The results of this study support the prediction that austenitic alloys irradiated at PWR- relevant displacement rates will most likely swell more than when irradiated at higher rates characteristic of fast reactors. Thus, the use of swelling data accumulated in fast reactors may possibly lead to an under-prediction of swelling in lower-flux PWRs and fusion devices. INTRODUCTION The void swelling phenomenon was the life-limiting aging phenomenon of austenitic stainless steels in fast reactors (1). Until very recently it was not thought that void swelling would occur in the austenitic internal components of light water reactors (LWRs). Similar predictions have been made for some fusion devices operating at lower dpa rates and lower temperatures. The two reactor concepts share many similarities, including high generation rates of helium and hydrogen. It was predicted in 1994, however, that the 316 and especially 304 stainless steels used in the baffle-former-barrel assembly of pressurized water reactors (PWRs) were probably already swelling at relatively low levels and would continue to swell at accelerating rate as the nuclear plants continued to age (2,3). Due to the accelerating development of swelling * Pacific Northwest National Laboratory (PNNL) is operated for the U.S. Department of Energy by Battelle Memorial Institute under contract DE-AC06-76RLO-1830.