1. 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

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 AlloyCrNiFeCMoCuWNbAlTiMnSi 304 SS18-208-11bal0.08 316 SS16-18 11-14 bal0.082-3 HT-9120.5bal0.201.00.50.60.4V:0.25 T-919bal0.101.00.080.450.4V:0.20, N:0.05 HCM12A 12bal0.110.41.02.00.050.60.1N:0.06, B:0.003 Alloy 60015.5768.00.080.50.2 Alloy 69030609.50.03 Alloy 62521.5612.50.059.03.60.2 Alloy 7181952.518.50.043.05.10.50.90.2 Austenitic stainless steels Solid solution Ni-base austenitic steels Precipitation hardened Ni-base austenitic steels Ferritic/martensitic steels Titanium alloys Ti-6Al-4V4 V6balSn,Cr,Zr,Mo

5. Austenitic stainless steels AlloyApplication historyProperties 304 SS LWR internals (BWR) Good corrosion resistance and moderated strength through the 400 – 500°C range. Loss of strength at the upper end of the range Susceptible to swelling Susceptible to SCC in 500°C SCW 316 SS LWR internals (PWR) Cladding in LMFBRs Better SCC resistance than 304 Localized corrosion (IGSCC) in oxidizing conditions and susceptibility to radiation-induced swelling in the 400-600°C range

6. Solid solution Ni-base austenitic alloy AlloyApplication historyProperties Alloy 600 Steam generator tubing in PWRs, control rod drive penetrations. Susceptible to IGSCC in PWR environment Susceptible to IG creep failure in any environment above 500°C Alloy 690 Replacement SG tubing in PWRs, control rod drive penetrations. Stronger and more corrosion resistant than Alloy 600 Less susceptible to IGSCC than Alloy 600 in PWR environment

7. Precipitation hardened Ni-base austenitic alloy AlloyApplication historyProperties Alloy 625 Reactor-core and control rod components in LWRs Hardened by  ” phase [Ni 3 (Nb,Al,Ti)] precipitated by aging. Higher strength up to about 700  C than solid solution Ni-base alloys. Susceptible to pitting in non-deaerated SCW at temperatures above 400°C Alloy 718 Reactor-core components Hardened by  ’ [Ni 3 (Ti, Al)] and  ” [Ni 3 (Nb,Al,Ti)] precipitated by aging. Higher strength up to about 700  C than solid solution Ni-base alloys. Susceptible to IGSCC in SCW High activation

8. Ferritic/martensitic steels AlloyComposition (wt%) Application historyProperties HT-9 Cr:12, Ni:0.5, Fe:Bal., C:0.2, Mo:1.0, Co:1.0, W:0.5, Mn:0.6, Si:0.4, V:0.25 Structural applications in supercritical fossil power plants Lower coefficient of thermal expansion and higher thermal conductivity than austenitic steels. T-91 Cr:9, Fe:Bal., C:0.1, Mo:1.0, Nb:0.08, Mn:0.45, Si:0.4, V:0.2, N:0.05 High swelling resistance High creep strength Higher corrosion rates than austenitic/Ni-base alloys Additions such as Cu may cause irrad. embrittlement HCM 12A Cr:12, Fe:Bal., C: 0.11, Mo:0.4, Cu:1.0, W:2.0, Nb: 0.05, Mn:0.6, Si:0.1, N:0.06, B:0.003

9. AlloyApplication historyProperties Ti-6Al- 4V Little application in commercial reactor systems Good corrosion resistance Good mechanical properties Uncertainty in radiation stability and in properties of irradiated state High cost 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.