Research and development progress is reported on fuel dissolution, solvent extraction studies, corrosion studies, mechanisms of foam separation, waste treatment, ion exchange, and chemical applications of nuclear explosions. (M.C.G.)

Report discussing a study of various neutron image detector systems and their characteristics, and opal glass diffusers in an optical image synthesis apparatus. A five inch diameter neutron beam exposure facility is also discussed.

The objective of this production test is to authorize the irradiation of three natural uranium corrugated I&E-type fuel elements in a KE Reactor front-to-rear test hole. The ultimate exposure of Zircaloy-2 clad, coextruded, uranium-cored fuel elements may be limited by localized tensile necking and splitting of the jacket. The total deformation capability of a fuel element is increased by corrugating the outside surface so the perimeter is greater than that of a circle circumscribing the equivalent cross sectional area. Thus, swelling of the uranium core is accommodated by bending of the clad rather than by large clad tensile strains. The test is designed so the knowledge gained will be applicable to the NPR fuel assemblies.

This report details the borehole examination and inspection results of process tube 2959 following a massive fuel failure and a low flow trip which scrammed the reactor.

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Twelve Zircaloy-2 clad UO/sub 2/ specimens, irradiated in the NRK to about 15,500 Mwd/t and decayed about 9 months, were declad and core dissolutions completed. The pellets were fractured brt were essentially intact (i.e. had not fallen apart) when declad. The U losses ramged from 0.03 to 0.09%, due to solubility of U(IV) in the Zinflex reagent, and the U losses from 0.01 to 0.04%. The core pellets were 99.5% dlssolved in 4 M HNO/sub 3/-0.1M Al(NO/sub 3/)/sub 3/ in 5 hr, which was slightly faster than the rate of unirradiated pellets. (auth)

Bare and zirconium-clad uranium-10 wt% molybdenum specimens were irradiated in NaK-filled capsules in the MTR. Irradiation conditions varied to include central-core temperatures ranging from 300 to over 1200 deg F, fuel burnups ranging from 0.36 to over 3.0 total at.% and fission rates in the range of 0.35 to 1.9 x 10/sup 14/ fissions/(sec)(cm/sup 3/) of alloy. Other parameters studied included the effects of heat treatment, changes in composition, different fabrication techniques, and changes in cladding thickness on the behavior of the fuel alloy. The objective of the irradiations was to determine the behavior of the fuel alloy under conditions approaching as closely as possible those to be encountered in the Enrico Fermi Atomic Power Plant as they were known at the time. The results indicated that the volume of the fuel alloy would increase conservatively at a rate of about 3.0% per at.% burnup as long as the critical temperature of 1000 to 1100 deg F was not exceeded and the gamma phase of the alloy did not transform during irradiation. If the critical temperature was exceeded, the alloy swelled until rupture or complete disintegration occurred. The occurrence of transformation during irradiation was noted at burnups in the range …