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Identification of most promising candidate alloys for fuel cladding and core internal structures

SCWR Information Meeting - April 29-30, 2003 UW-Madison Materials and Chemistry. Identification of most promising candidate alloys for fuel cladding and core internal structures . SCWR Environment. Light water at 25 MPa Temperature range: 280-600°C

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Identification of most promising candidate alloys for fuel cladding and core internal structures

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  1. SCWR Information Meeting - April 29-30, 2003 UW-Madison Materials and Chemistry Identification of most promising candidate alloys for fuel cladding and core internal structures

  2. SCWR Environment Light water at 25 MPa Temperature range: 280-600°C Coolant chemistry: unknown, but likely to contain dissolved oxygen in the hundred ppb range. Components • Fuel cladding, spacer grids/wire wrap, water rod boxes, ducts • Lower core plate, upper support plate, CR guide tubes • Core barrel • Pressure vessel • Ex-core components - “steam” lines, turbine components, etc.

  3. Clad and Structural Materials Requirements • High temperature strength • yield strength, ductility, creep • Corrosion • uniform • localized • stress corrosion cracking • Radiation stability • RIS, microstructure, voids/swelling, creep, growth, phase • Irradiated state properties • strength, ductility, creep, corrosion/SCC, fracture toughness, fatigue

  4. Candidate alloy systems Austenitic stainless steels Solid solution Ni-base austenitic steels Precipitation hardened Ni-base austenitic steels Ferritic/martensitic steels Titanium alloys

  5. Austenitic stainless steels

  6. Solid solution Ni-base austenitic alloy

  7. Precipitation hardened Ni-base austenitic alloy

  8. Ferritic/martensitic steels

  9. Titanium Alloys

  10. Candidate Alloy Systems 304 SS and 316 SS provide links to extensive database in BWR and PWR conditions to evaluate the cracking behavior across an extended range of temperature and environment. Nickel-base should have better corrosion resistance and high temperature strength, but are likely to be susceptible to pitting and IGSCC. Ferritic-martensitic alloys are most promising but have no history in reactor systems. Titanium alloys are big unknowns. None of the systems has the benefit of property data in SCW conditions in the irradiated state.

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