Report of the test results of a pulsed eddy-current apparatus for inspecting the integrity of smooth and ribbed tubing as well as discussion of transducer and preamplifier features, scanning procedures, and theory of operation.

A nondestructive autoradiographic method is described which can provide a verification that rods in the interior of unirradiated LWR fuel assemblies contain low-enriched uranium. Sufficient absorber must be used to reduce contributions to image density by beta radiation from uranium-238 daughters. When appropriate absorbers are used, the density of the image of a uranium-containing fuel rod is proportional to the uranium-235 enrichment in that rod. Exposure times as short as 1.5 hours can be achieved by using fast film and intensifying screens. Methods are discussed for reducing contributions to the image density of any single rod from radiation produced by all other rods in the assembly. The technique is useful for detecting missing rods, dummy rods, and rods containing depleted uranium. These defects can be detected by visual inspection of the autoradiographs. In its present state of development, the technique is not sensitive enough to reliably detect the difference between the various uranium-235 enrichments encountered in current BWR fuel assemblies. Results are presented for field tests of the technique at BWR and PWR facilities.

Subassembly XX08 is a fueled and instrumented subassembly designed primarily for an ongoing program to investigate the thermal-hydraulic core environment within EBR-II under normal and off-normal plant operating conditions. XX08 contains 58-xenon-tagged, EBR-II Mark-II driver-fuel elements. The Mark-II fuel is expected to provide XX08 with an irradiation lifetime three times as great as that attained with its predecessor, XX07, i.e., a 9 versus 2.9% burnup. A burnup of 9 at.% is equivalent to about 29,000 MWt dyays of EBR-II reactor operation, which corresponds to 11 reactor runs at 2700 MWd per run.

TREAT F-series tests are being conducted to provide data on fuel motion in an LMFBR during a hypothetical loss-of-flow accident. Fuel and fuel-boundary conditions in an LMFBR subassembly following sodium voiding and dryout under loss-of-flow conditions are simulated in each F-series test. Simulation is achieved with a single fuel element surrounded by an annular nuclear-heated wall in a dry (no sodium) test capsule. The area inside the heated wall was selected to represent the area inside the perimeter of an LMFBR coolant channel. Test F1 was conducted with an irradiated fuel element to investigate the effect of fission gas on fuel motion at design power levels following cladding melting and drainage. The principal conclusion from Test F1 is that fission products retarded, but did not prevent, eventual fuel collapse. The collapse was retarded by a fuel/fission-product froth that prevented fuel collapse until the fission products separated from the partially molten fuel. The fuel motion observed in F1 represents a particular type of fuel (burnup of 2.35 at.%, power rating of 394 W/cm) transient heated at design power rating.

Postirradiation examinations were performed on five fuel pins from the Gas-Cooled Fast-Breeder Reactor F-1 experiment irradiated in EBR-II to a peak burnup of approximately 5.5 at. %. These encapsulated fuel pins were irradiated at peak-power linear ratings from approximately 13 to 15 kW/ft and peak cladding inside diameter temperatures from approximately 625 to 760°C. The maximum diametral change that occurred during irradiation was 0.2% .delta.D/D₀. The maximum fuel-cladding chemical interaction depth was 2.6 mils in fuel pin G-1 and 1 mil or less in the other three pins examined destructively. Significant migration of the volatile fission products occurred axially to the fuel-blanket interfaces. The postirradiation examination data indicate that fuel melted at the inner surface of the annular fuel pellets in the two highest power rating fuel pins, but little axial movement of fuel occurred.

A reference conceptual flowsheet based on existing or developing technology for encapsulation of stabilized calcine pellets is discussed. Conclusions and recommendations are presented.