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Resources for deep understanding about nuclear fuel materials

  1. Oct 8, 2009 #1
    Hello,everybody!
    I'm a PhD candidate from xi'an jiaotong university,china. There is an urgent need to make a deep understanding about the microstructure evolution of nuclear fuel materials under irradiation. To make my research more close to the frontiers, your suggestions are welcomed. I've devoted myself to the work independently, but until now, I can't find a good perspective to this field. So,my primary questions follow as,
    1.what is the specific characteristic of the microstructure evolution of nuclear fuel materials?
    2.what method can accurately analyze and predict the fuel behavior under high burnup?
    3.Are there any communities or anybody else concentrating on this subject? Is it possible to communicate some ideas with each other?
    Thanks a lot!
    Best regards!
    waiting for your help!
     
  2. jcsd
  3. Oct 8, 2009 #2

    Astronuc

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    Staff: Mentor

    One of my primary areas is nuclear fuel and core technology, and materials used therein.

    Note that there are different materials which provide different functions.

    The fuel proper is the fissile elements U-233, U-235, and Pu-239, 241 which are usually embedded in U-238 or inert matrix, which are usually ceramic forms. The fission process produces fission products that accumulate with burnup, so understanding fission product behavior (of gases, volatiles, non-metals and metals) is critical to understanding fuel behavior with burnup. The fuel matrix is usually UO2 or (U,Pu)O2, but it could be MC (carbide) or MN (nitride), where M = U and/or Pu. Carbides and nitrides have higher density and thermal conductivity, but are usually not used in LWR because of the reaction with water. Some fuel designs use ceramic-metal (cermet) matrices, and still other use metal or metal hydride matrices. The behaviors of these forms have similarities, but are also quite different as a function of burnup.

    Surrounding the fuel material is the cladding, which in LWRs is usually a Zr-alloy (e.g., Zr-2 in BWRs or Zr-4 in PWRs, or more recently Zr-Nb alloys such as AREVA's M5 or Westinghouse's ZIRLO). The key issues with these alloys are corrosion/oxidation as a function of duty and burnup, irradiation hardening (with the consequence of 'notch sensitivity') and hydrogen pickup, which is a consequence of the oxidation reaction between Zr and H2O. Some people are studying the use of ceramic claddings.

    For fast or liquid metal cooled reactors, the cladding material has traditionally been austenitic (e.g. SS316) or ferritic stainless steels.

    In addition to changes in mechanical properties with irradiation, we are concerned about dimensional stability of the fuel and core materials.

    As for community(ies), there are now annual meeting held in a rotational basis in US, Europe and Asia. The European Meeting is named TOPFUEL, and in the US it is the International Topical Meeting on Light Water Reactor Fuel Performance. There are also embedded topical meetings every two years in the ANS (American Nuclear Society) Summer Meeting. Within ANS, one will find the Material Science and Technology Division (MSTD). Many of the participants work at the fuel vendors AREVA, GNF/GEH, Toshiba/Westinghouse, the DOE labs, and academia.

    One will find work published in the Journal of Nuclear Materials, which is published by Elsevier. ASTM has a triennial meeting or International Symposium: Zirconium in the Nuclear Industry.
     
  4. Oct 8, 2009 #3
    Ceramic cladding? I haven't heard of that before, but it seems like metal is more advantageous concerning heat transfer and neutron capture.
     
  5. Oct 8, 2009 #4

    Astronuc

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    Staff: Mentor

    SiC for LWR cladding - believe it or not. Personally, I'm skeptical.
     
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