Post-irradiation Annealing of Dislocation Microstructure and ...

Post-irradiation Annealing of Dislocation Microstructure and ...

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: 280-600C 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. 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

Candidate alloy systems Alloy Cr Ni Fe C Mo Cu W Nb Al Ti Mn Si 0.5 0.2 Austenitic stainless steels 304 SS 18-20 8-11 bal

0.08 316 SS 16-18 11-14 bal 0.08 2-3 Solid solution Ni-base austenitic steels Alloy 600 15.5 76 8.0 0.08 Alloy 690 30 60 9.5 0.03 Precipitation hardened Ni-base austenitic steels Alloy 625 21.5 61 2.5 0.05

9.0 3.6 0.2 0.2 0.2 0.2 Alloy 718 19 52.5 18.5 0.04 3.0 5.1 0.5 0.9 0.2 0.2 bal 0.20 1.0

0.6 0.4 V:0.25 0.08 0.45 0.4 V:0.20, N:0.05 0.05 0.6 0.1 N:0.06, B:0.003 Ferritic/martensitic steels HT-9 12 T-91 9 bal 0.10 1.0 HCM12A

12 bal 0.11 0.4 Titanium alloys Ti-6Al-4V 0.5 0.5 1.0 2.0 4V 6 bal Sn,Cr,Zr,Mo Austenitic stainless steels Alloy Application history Properties 304 SS LWR internals (BWR) Good corrosion resistance and

moderated strength through the 400 500C range. Loss of strength at the upper end of the range Susceptible to swelling Susceptible to SCC in 500C SCW 316 SS LWR internals Better SCC resistance than 304 (PWR) Localized corrosion (IGSCC) in Cladding in LMFBRs oxidizing conditions and susceptibility to radiation-induced swelling in the 400-600C range Solid solution Ni-base austenitic alloy Alloy Application history Properties 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 500C 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 Precipitation hardened Ni-base austenitic alloy Alloy Application history Properties Alloy 625 Reactor-core and Hardened by phase [Ni3(Nb,Al,Ti)] control rod components precipitated by aging. in LWRs Higher strength up to about 700C than solid solution Ni-base alloys. Susceptible to pitting in non-deaerated SCW at temperatures above 400C Alloy 718 Reactor-core components Hardened by [Ni3(Ti, Al)] and [Ni3(Nb,Al,Ti)] precipitated by aging. Higher strength up to about 700C than solid solution Ni-base alloys. Susceptible to IGSCC in SCW High activation Ferritic/martensitic steels Alloy Composition

(wt%) 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 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 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 Application history Properties Structural applications Lower coefficient of in supercritical fossil thermal expansion and power plants higher thermal conductivity than austenitic steels. High swelling resistance High creep strength Higher corrosion rates than austenitic/Ni-base alloys

Additions such as Cu may cause irrad. embrittlement Titanium Alloys Alloy Application history Properties Ti-6Al4V Little application in commercial reactor systems Good corrosion resistance Good mechanical properties Uncertainty in radiation stability and in properties of irradiated state High cost 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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