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Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U–Zr fuel is a driver for the initial core of the PGSFR, and U–transuranics (TRU)–Zr fuel will gradually replace U–Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U–Zr fuel, work on U–Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U–TRU–Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic–martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

Candidate coating materials for re-usable metallic nuclear fuel crucibles, TaC, TiC, ZrC, ZrO2, and Y2O3, were plasma-sprayed onto a niobium substrate. The microstructure of the plasma-sprayed coatings and thermal cycling behavior were characterized, and U-Zr melt interaction studies were carried out. The TaC and Y2O3 coating layers had a uniform thickness, and high density with only a few small closed pores showing good consolidation, while the ZrC, TiC, and ZrO2 coatings were not well consolidated with a considerable amount of porosity. Thermal cycling tests showed that the adhesion of the TiC, ZrC, and ZrO2 coating layers with niobium was relatively weak compared to the TaC and Y2O3 coatings. The TaC and Y2O3 coatings had better cycling characteristics with no interconnected cracks. In the interaction studies, ZrC and ZrO2 coated rods showed significant degradations after exposure to U-10 wt.% Zr melt at 1600°C for 15 min., but TaC, TiC, and Y2O3 coatings showed good compatibility with U-Zr melt.

The fabrication technology for metallic fuel has been developed to produce the driver fuel in a PGSFR in Korea since 2007. In order to evaluate the irradiation integrity and validate the in-reactor of the starting metallic fuel with FMS cladding for the loading of the metallic fuel, U-10 wt.%Zr fuel rodlets were fabricated and evaluated for a verification of the starting driver fuel through an irradiation test in the BOR-60 fast reactor. The injection casting method was applied to U-10 wt.%Zr fuel slugs with a diameter of 5.5 mm. Consequently, fuel slugs per melting batch without casting defects were fabricated through the development of advanced casting technology and evaluation tests. The optimal GTAW welding conditions were also established through a number of experiments. In addition, a qualification test was carried out to prove the weld quality of the end plug welding of the metallic fuel rodlets. The wire wrapping of metallic fuel rodlets was successfully accomplished for the irradiation test. Thus, PGSFR fuel rodlets have been soundly fabricated for the irradiation test in a BOR-60 fast reactor.

U-10wt%Zr spherical particles for use as particulate fuel were prepared by centrifugal atomization and subjected to pressureless sintering, which is one of the simplest powder processing techniques. At sintering temperature of 1100°C for 30 or 60 min, all samples ranging from +50 to −325 mesh showed no apparent bonding between the particles. However, at 1150°C (80 min), all samples formed a bulk body and the microstructures showed apparent sintering stages. Particularly, sample B (50–70 mesh) and sample C (70–100 mesh) showed pore characteristics suitable for a particulate fuel. The results suggest that pressureless sinterability for U-10Zr particulate fuel can be improved by adding small-size (–325 mesh) particles.

Microstructural and mechanical property were evaluated at the medium-sized HT9 (12Cr-1MoWV) forged steel which was considered as primary candidate for the fuel cladding in sodium-cooled fast reactor (SFR). Material was forged at after the induction melting to make round bar as 160mm diameter, 7000mm length then the radial distribution of microstructure as well as microhardness was evaluated. The results showed that overall microstructure exhibited as ferrite-martensite structure, where small amount (2~3%) of delta ferrite was formed throughout the specimen and maximum 15% of transformed ferrite was formed at the center, where it gradually decreased toward the radial direction. Sensitivity analysis of the cooling curve and Time-Temperature-Transformation (TTT) diagram revealed that formation of transformed ferrite could be avoided when the diameter was decreased down to 120mm.

This study aimed to evaluate the performance of diffusion barriers in order to prevent fuel-cladding chemical interaction (FCCI) between the metallic fuels and the cladding materials, a potential hazard for nuclear fuel in sodium-cooled fast reactors. In order to prevent FCCI, Zr or V metal is deposited on the ferriticmartensitic stainless steel surface by physical vapor deposition with a thickness up to 5μm. The diffusion couple tests using uranium alloy (U-10Zr) and a rare earth metal such as Ce-La alloy and Nd were performed at temperatures between 660~800°C. Microstructural analysis using SEM was carried out over the coupled specimen. The results show that significant interdiffusion and an associated eutectic reaction ocurred in the specimen without a diffusion barrier. However, with the exception of the local dissolution of the Zr layer in the Ce-La alloy, the specimens deposited with Zr and V exhibited superior eutectic resistance to the uranium alloy and rare earth metal. (Received January 17, 2011)

In order to meet the increasing demand for electricity, Korea has to rely on nuclear energy due to its poor natural resources. In order for nuclear energy to be expanded in its utilization, issues with uranium supply and waste management issues have to be addressed. Fast reactor system is one of the most promising options for electricity generation with its efficient utilization of uranium resources and reduction of radioactive waste, thus contributing to sustainable development. The Korea Atomic Energy Research Institute (KAERI) has been performing R&Ds on Sodium-cooled Fast Reactors (SFRs) under the national nuclear R&D program. Based on the experiences gained from the development of KALIMER conceptual designs of a pool-type U-TRU-10%Zr metal fuel loaded reactor, KAERI is currently developing Advanced SFR design concepts that can better meet the Generation IV technology goals. This also includes developing, Advanced SFR technologies necessary for its commercialization and basic key technologies, aiming at the conceptual design of an Advanced SFR by 2011. KAERI is making R&D efforts to develop advanced design concepts including a passive decay heat removal system and a supercritical CO 2 Brayton cycle energy conversion system, as well as developing design methodologies, computational tools, and sodium technology. The long-term Advanced SFR development plan will be carried out toward the construction of an Advanced SFR demonstration plant by 2028.

Based on a composition of 99.4 wt% AISI 316L stainless steel, 0.3 wt% Ti and 0.3 wt% Y2O3, an austenitic ODS steel was fabricated by a process of mechanical alloying, hot isostatic pressing and rolling. Fine oxide particles were observed in the matrix, and their chemical formulations were determined to be Y2Si2O7 and TiO. Heat treatment of the cold-rolled sample at 1200°C induced an isotropic tensile behavior at room temperature and at 700°C This result would be mainly attributed to the equiaxed grains that form as a result of the heat treatment for recrystallization.

The Korea Atomic Energy Research Institute has developed an advanced fast reactor concept, KALIMER-600, which satisfies the Generation IV reactor design goals of sustainability, economics, safety, and proliferation resistance. The concept enables an efficient utilization of uranium resources and a reduction of the radioactive waste. The core design has been developed with a strong emphasis on proliferation resistance by adopting a single enrichment fuel without blanket assemblies. In addition, a passive residual heat removal system, shortened intermediate heat-transport system piping and seismic isolation have been realized in the reactor system design as enhancements to its safety and economics. The inherent safety characteristics of the KALIMER-600 design have been confirmed by a safety analysis of its bounding events. Research on important thermal-hydraulic phenomena and sensing technologies were performed to support the design study. The integrity of the reactor head against creep fatigue was confirmed using a CFD method, and a model for density-wave instability in a helical-coiled steam generator was developed. Gas entrainment on an agitating pool surface was investigated and an experimental correlation on a critical entrainment condition was obtained. An experimental study on sodium-water reactions was also performed to validate the developed SELPSTA code, which predicts the data accurately. An acoustic leak detection method utilizing a neural network and signal processing units were developed and applied successfully for the detection of a signal up to a noise level of -20 dB. Waveguide sensor visualization technology is being developed to inspect the reactor internals and fuel subassemblies. These research and developmental efforts contribute significantly to enhance the safety, economics, and efficiency of the KALIMER-600 design concept.

Metallic fuel slugs of U–10Zr–5Mn (wt%), a surrogate alloy for the U–TRU–Zr (TRU: a transuranic element) alloys proposed for sodium-cooled fast reactors, were prepared by injection casting in a laboratory-scale furnace, and their characteristics were evaluated. As-cast U–Zr–Mn fuel rods were generally sound, without cracks or thin sections. Approximately 68% of the original Mn content was lost under dynamic vacuum and the resulting slug was denser than those prepared under Ar pressure. The concentration of volatile Mn was as per the target composition along the entire length of the rods prepared under 400 and 600 Torr. Impurities, namely, oxygen, carbon, silicon, and nitrogen, totaled less than 2,000 ppm, satisfying fuel criteria.

A study has been carried out to evaluate the interaction behavior between a lanthanide element and clad material in order to analyze the effect of the lanthanide element on the fuel cladding chemical interaction (FCCI). A diffusion couple test between Misch metal (70Ce-30La) and ferritic-martensitic steel (Gr.92) was performed at 660 degrees C, followed by a microstructural analysis of the coupled sample. The results showed that Ce in the Misch metal, rather than La, reacted with the ferritic-martensitic steel (FMS) to form an interaction layer that penetrated the clad thickness. Fe diffused outside the clad interface to form an Fe(2)Ce compound, leaving a depletion of Fe caused by excess diffusion as well as by the formation of Cr-rich precipitation inside the interaction layer. The rate of growth followed the cubic rate law, which indicated that Fe depletion was caused by the diffusion of Fe and that the associated Cr-rich phase formation controlled the whole diffusion process.