Experimental studies of U-Pu-Zr fast reactor fuel pins in the experimental breeder reactor-ll
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INTRODUCTION
T H E IFR is a generic reactor concept under development by Argonne National Laboratory. Much of the technology for the IFR is being demonstrated at the EBRII on the Department of Energy's site near Idaho Falls, ID. The IFR concept relies on four technical features to achieve breakthroughs in nuclear power economics and safety: (1) a pool-type reactor configuration, (2) liquid sodium cooling, (3) metallic fuel, and (4) an integral fuel cycle with on-site reprocessing, tl] The purpose of this paper will be to summarize the latest results of irradiation testing of uranium-plutonium-zirconium (U-Pu-Zr) fuel in the EBR-II. II.
BACKGROUND
In order to better appreciate the materials aspects of these fuel tests, a brief description of the key design features will be given. The ultimate criteria for fuel pin design is the overall integrity up to the planned lifetime or target burnup of irradiation. Here, burnup refers to the fraction of total heavy metal constituents (U + Pu) which have undergone fission and, hence, transferred thermal power to the sodium reactor coolant medium flowing around the pins. Achievement of burnups near 15 at. pct is desirable for large-scale commercialization of liquidmetal reactors (LMR). Figure 1 illustrates the key design features of a generic metallic fuel pin along with some typical specifications for actual IFR test pins irradiated R.G. PAHL, Metallurgical Engineer and Technical Group Leader, C.E. LAHM, Mechanical Engineer, and D.L. PORTER, Metallurgist and Manager, Reactor Materials Section, are with the Experimental Breeder Reactor-l] Division, Argonne National Laboratory, Idaho Falls, ID 83403. G.L. HOFMAN, Senior Metallurgist, is with Argonne National Laboratory, Argonne, IL 60439. This paper is based on a presentation made in the symposium "Irradiation-Enhanced Materials Science and Engineering" presented as part of the ASM INTERNATIONAL 75th Anniversary celebration at the 1988 World Materials Congress in Chicago, IL, September 25-29, 1988, under the auspices of the Nuclear Materials Committee of TMS-AIME and ASM-MSD. METALLURGICAL TRANSACTIONS A
in this study. The fuel itself is a solid cylinder which has been injection-cast into quartz molds and loaded into the cladding jacket with no thermomechanical pretreatment required. The as-cast density of these alloys is ~ 15.8 g / cm 3. Sodium fills the annulus around the slug and provides a high conductivity heat path to the cladding prior to fuel slug swelling. The plenum space (initially filled with inert gas) above the fuel slug is provided to contain the large volume of stable fission gas (Xe/Kr) which is produced in these fuel slugs at the rate of --16 cm3 (STP) per percent burnup. The ends of the jacket are hermetically sealed with an automated tungsten inert gas welder, and a helical wire is welded in place to maintain pin spacing and mix coolant flow upward through the closepacked hexagonal pin bundle during irradiation. The key phenomena which are significant in controlling fuel pin behavior and reliability ar
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