[en] This paper constitutes an interim report on the sodium-water reaction programs performed, using the GE-SOWAT, GE-SMALL LEAK BEHAVIOR RIG, and GE-PTTR facilities in support of LMFBR steam generator development and its application to the Clinch River Breeder Reactor Plant. Test data from these rigs are presented, including wastage data as a function of water injection rate, sodium temperature, and orifice geometry. Initial results for self-wastage of defects under prototypical conditions, and from proof-of-principle tests of a protected heat transfer tube concept are also presented. An analytical basis for wastage phenomena is suggested. (author)