Development of Eutectic Free Cladding Materials for Metallic Fuel (original) (raw)
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CLADDING SURVEY FOR THE ENRICO FERMI REACTOR U-15 Wt.% Mo BASE DISPERSION- TYPE FUEL ELEMENT
1960
Potential cladding materials for a flat-plate fuel element containing a dispersion of UC or U0 2 in U-15 wt $ Mo alloy were surveyed on the bases of'compatibility with the fissile compounds, matrix material, protective cover materials, and liquid .sodium as well .as the feasibility of fabricating fuel plates by roll cladding. Radiative-capture cross sections, thermodynamic data, eutectic and intermediate compound formation, mechanical properties, and corrosion by 1000°F sodium were reported for austenitic stainless steels, chromium, nickel, niobium, molybdenum, tantalum, vanadium, and zirconium. It was recommended that "A" nickel (molybdenum barrier), Zr-3 wt Al, Nb-2 wt $ Cr, and Fansteel 82 be selected for investigation. NOTICE This document contains information of a preliminary nature and was prepared primarily for internal use at the Oak Ridge National Laboratory. It is sub|ect to revision or correction and therefore does not represent a final report. The information is not to be abstracted, reprinted or otherwise given public dissemination without the approval of the ORNL patent branch, Legal and Information Control Department.
Zirconium Alloys for Fuel Element Structures
CHIMIA, 2005
Today more than 400 light water power reactors (LWRs) operate worldwide providing approximately 17% of the world's electricity demand. One important component for their successful operation is the fuel tube, made out of a zirconium alloy. A huge number of out-of-pile and in-pile experiments have been performed to improve step by step the fuel for higher burn-up and to reduce the failure rates of fuel pins close to zero. The influencing parameters for excellent or poor cladding behaviour are numerous and sometimes counteract each other. The process of cladding corrosion is slow, difficult to follow, the mechanistic understanding at best incomplete. A vast amount of literature documents the abundant tests and comes up with hypotheses and models for the materials behaviour. PSI has supported for the past 20 years the development of high burn-up fuel cladding by microstructural research studies and service work in post-irradiation examination of test pins. This article reviews the d...
Performance of FCCI barrier foils for U–Zr–X metallic fuel
Journal of Nuclear Materials, 2009
Diffusion couple tests of U-Zr or U-Zr-Ce alloys vs. ferritic martensitic steels such as HT9 or T91 were carried out in order to evaluate the performance of the diffusion barrier candidates. Elemental metal foils of Zr, Nb, Ti, Mo, Ta, V and Cr were very effective in inhibiting interdiffusion between these fuels and steels. Eutectic melting between the fuels and steels was not observed in any of the diffusion couples using these diffusion barrier foils at annealing temperatures up to 800°C. Among the metallic foils evaluated in this study, V and Cr exhibited the most promising performances as a diffusion barrier material for eliminating the fuel cladding chemical interaction problem. However, Zr, Nb and Ti showed an active interaction with the fuel mainly due to the large U solubility.
Nuclear Science and Engineering, 2020
Alloying additions are introduced into U-Zr fuel in order to bind lanthanides (e.g., cerium) and prevent their migration to the fuel-cladding interface. Antimony (Sb) is being investigated as a candidate additive. The present study focuses on the diffusion couple behavior of U-10Zr (wt%) alloy with Sb against cladding (iron or HT9) at 640°C. The diffusion cross sections were analyzed using a scanning electron microscope and X-ray diffraction. Zr-rind was found at the interface of the fuel alloy, Sb was found to be bound in Sb-Zr precipitates or Sb-Ce precipitates, and no reaction was found between Sb precipitates and the cladding materials.
Fuel clad chemical interaction of U-Mo fast reactor fuel
Journal of Nuclear Materials, 2019
U-33 at.% Mo (16.8 wt.%) metallic fuel is a candidate material for metallic fuel Fast Reactor. One of the lifelimiting factors of fast breeder reactor clad is the fuel clad chemical interaction due to formation of low melting eutectics between U and Fe. This chemical interaction should be avoided or minimized to increase the fuel burn-up. Fuel clad chemical interaction between U-33 at.% Mo metallic fuel with T91 (9 Cr-1Mo) clad has been studied at 650 C, 675 C and 700 C, for different soaking time, through diffusion couple experiments. Development of microstructures, phase constituents and compositions during thermal treatment were examined by scanning electron microscopy and X-ray energy dispersive spectroscopy. In the fuel side, U 6 Fe phase is formed along with bcc-Mo in lamellar morphology through cellular precipitation. Due to slower diffusion of Mo compared to uranium from U-33 at.% Mo fuel, a Mo rich layer is formed at the slug surface. This Mo rich layer subsequently acts as a diffusion barrier layer and minimizes further growth of clad-wastage zone and slug-interaction zone. Multi-phase layer growth constants and activation energies have been calculated and compared with interactions reported in literature of other U-based fuels and cladding elements. The study indicates that U-33 at.% Mo/T91 fuel reduces fuel clad chemical interaction significantly compared with U-23Zr/Fe and U-23Zr/Fe-Cr.
Metal Fuel Development and Verification for Prototype Generation IV Sodium-Cooled Fast Reactor
Nuclear Engineering and Technology, 2016
Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. UeZr fuel is a driver for the initial core of the PGSFR, and U etransuranics (TRU)eZr fuel will gradually replace UeZr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of UeZr fuel, work on UeZr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. UeTRU eZr 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 ferriticemartensitic 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.
2013
Zirconium-based alloys Zircaloy-2 and Zircaloy-4 are widely used in the nuclear industry as cladding materials for light water reactor (LWR) fuels. These materials display a very good combination of properties such as low neutron absorption, creep behavior, stress-corrosion cracking resistance, reduced hydrogen uptake, corrosion and/or oxidation, especially in the case of Zircaloy-4. However, over the last couple of years energetic efforts have been undertaken to improve fuel clad oxidation resistance during off-normal temperature excursions. Efforts have also been made to improve upon the already achieved levels of mechanical behavior and reduce hydrogen uptake. In order to facilitate the development of such novel materials, it is very important to achieve not only engineering control, but also a scientific understanding of the underlying material degradation mechanisms, both in working conditions and in storage of used nuclear fuel.
Corrosion Behavior of Zirconium Alloy Nuclear Fuel Cladding
MRS Proceedings, 1989
ABSTRACTZircaloy−2 and −4 are used as nuclear fuel cladding. Both alloys are more than ninety-eight percent zirconium and are corrosion resistant to various media. Electrochemical measurements using polarization techniques have been made on these alloys in aqueous media with a pH of 8.5 and varying ionic concentration (1X and 10X) at temperatures of 22°C and 95°C. Results showed that under the test conditions of the study these alloys passivated and had negligible corrosion rates, but there were some variations in passivation due to surface preparation and some crevice corrosion was observed. Data are presented and discussed in terms of passivity, breakdown potential and susceptibility to localized corrosion.