[en] In-reactor, safety experiments performed in the Sodium Loop Safety Facility (SLSF) rely on comprehensive posttest examinations (PTE) to characterize the postirradiation condition of the cladding, fuel, and other test-subassembly components. PTE information and on-line instrumentation data, are analyzed to identify the sequence of events and the severity of the accident for each experiment. Following in-reactor experimentation, the SLSF loop and test assembly are transported to the Hot Fuel Examination Facility (HFEF) for initial disassembly. Goals of the HFEF-phase of the PTE are to retrieve the fuel bundle by dismantling the loop and withdrawing the test assembly, to assess the macro-condition of the fuel bundle by nondestructive examination techniques, and to prepare the fuel bundle for shipment to the Alpha-Gamma Hot Cell Facility (AGHCF) at Argonne National Laboratory

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[en] The thermionic technology program addresses the feasibility issues of a seven-year-life thermionic fuel element (TFE) for the SP-100 Thermionic Reactor Space Power System. These issues relate to the extension of TFE lifetime from three to seven years, one of the SP-100 requirements. The technology to support three-year lifetimes was demonstrated in the earlier TFE development program conducted in the late-1960s and 1970s. Primary life-limiting factors were recognized to be thermionic emitter dimensional increases due to swelling of the nuclear fuel and electrical structural damage from fast neutrons. The 1984-85 technology program is investigating the fueled emitter and insulator lifetime issues, both experimentally and analytically. The goal is to analytically project the lifetime of the fueled emitter and insulator and to experimentally verify these projection methods. In 1984, the efforts were largely devoted to the design and building of fueled emitters for irradiation in 1985, validation of fuel-emitter models, development of irradiation-resistant metal-ceramic seal and sheath insulator, modeling of insulator lifetime, and development of wide-spread, high-performance thermionic converters

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[en] The axial location of cladding failure following a transient overpower accident is of importance in fast reactor safety studies in that it is a determining factor in the relocation of fuel, and therefore in the possibility of inherent neutronic shutdown of the reactor. In-pile experimental data on the axial location of cladding failure of fuel in bundles of pins is sparse since, in general, the experimental fuel pin bundles are largely destroyed during the in-pile test. The post-test examination work has been completed for TREAT test J1. It was found that damage to the fuel elements during the irradiation was low enough for an accurate observation of the location of cladding failure to be made for each of the seven pins

[en] A series of CT codes is under development in the Reactor Analysis and Safety Division of Argonne National Laboratory for use as a post-test examination tool to analyze segments of the final fuel-bundle configuration of TREAT tests. This paper presents the results of CT analysis for fuel assemblies using neutron radiography. Fuel relocation following overpower transients in the TREAT reactor is examined for sections of the assemblies, and results are compared to metallographic sections. Further improvements are expected to increase the use and reliability of CT analysis as a standard post-test examination tool

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[en] Reliable knowledge of thermal conductivity is needed to assess reactor fuel performance. In normal operation, fuel pin linear power and fuel thermal conductivity together largely determine the fuel's peak temperature, and under severe accident conditions, fuel thermal conductivity largely determines the power to melt. Unfortunately, assessing the thermal conductivity of modern high-swelling sodium-bonded metallic fuel is neither theoretically straightforward nor amenable to laboratory measurements. At issue is whether the high swelling that takes place in metallic fuel during the first few atomic percent of burnup could lead to drastic conductivity reductions by factors of as much as 0.4 to 0.5. Such reduction may be mitigated by the infiltration of fuel porosity by high-conductivity liquid bond sodium or later by the solid and liquid fission products that accumulate with burnup. In this paper, the authors utilize measurements of maximum melting from intact irradiated metal fuel pins tested in the recent M-series in-pile test program in TREAT to estimate metal fuel thermal conductivity directly and in situ. Of six irradiated integral fast reactor prototype fuel pins ubjected to severe overpower and melting, four pins remained intact. Three of the test pins were of ternary alloy (U-19 Pu-10 Zr) and one of binary (U-10 Zr) alloy. Importantly, these test pins span key range of low burnup where minimum values of conductivity would be expected

[en] Transient Reactor Test (TREAT) Facility experiment M2 was performed to evaluate the transient behavior of metal-alloy fuel under accident conditions to investigate the inherent safety features of the fuel in integral fast reactor (IFR) system designs. Objectives were to obtain early information on the key fuel behavior characteristics at transient overpower (TOP) conditions in metal-fueled fast reactors; namely, margin to cladding breach and extent of axial self-extrusion of fuel within intact cladding. The onset of cladding breaching depends on fuel/cladding eutectic formation, as well as cladding pressurization and melting. Driving forces for fuel extrusion are fission gas, liquid sodium, and volatile fission products trapped within the fuel matrix. The post-test examination provided data essential for correctly modeling fuel behavior in accident codes