Gary Was

Professor

gsw@umich.edu

1921 Cooley

T: (734) 763-4675

Bio

Projects

Publications

Facilities

Group


Stress Corrosion Cracking and Corrosion of Candidate Alloys for the Supercritical Water Reactor Concept

Sponsor: DOE-NERI program
Supercritical water presents unique challenges to the long-term operation of engineering materials. The generation of oxygen and hydrogen gas by radiolysis and the high solubility of these gases in supercritical water may result in higher corrosion and stress corrosion cracking rates than experienced with other reactor designs. In addition, radiation may accelerate or assist the stress corrosion cracking in the reactor region, and stress corrosion cracking and accelerated corrosion may occur in the preheat and cool-down sections of the circuit. The existing data base on the corrosion and stress corrosion cracking of austenitic stainless steel and nickel based alloys in supercritical water is very sparse. Data on the behavior of irradiated alloys is non-existent. Therefore, the focus of this work will be stress-corrosion-cracking behavior of candidate fuel cladding and structural materials in the unirradiated and irradiated conditions. Two high-temperature autoclave systems have been built to test the SCC and corrosion behavior of unirradiated and proton-irradiated materials. Proton irradiation is used as a surrogate for neutron irradiated material in order to get a first look at the response of candidate alloys to irradiation, and also to cover alloys for which there are currently no neutron irradiated samples for testing. A third high-temperature autoclave coupled to a loading system, and capable of straining up to 4 tensile samples in constant extension rate mode or one compact tension sample in crack growth rate mode is being built and operated at the University of Michigan (UM). This system is being constructed for conducting experiments on neutron-irradiated materials. The resulting data will be used to further narrow the list of promising materials and develop appropriate stress-corrosion-cracking correlations. The capability to conduct both crack growth rate and constant extension rate tensile experiments on neutron-irradiated samples will constitute the first facility capable of assessing SCC of neutron irradiated alloys in supercritical water.The work plan for this three year (FY05-FY07) program consists of four principal tasks; 1) the completion of a facility to conduct crack growth rate and constant extension rate tensile tests on highly radioactive, neutron irradiated samples in supercritical water, 2) constant extension rate tests and crack growth rate tests of candidate alloys in supercritical water, 3) proton irradiation and constant extension rate tests of proton-irradiated samples in supercritical water and 4) constant extension rate tests and crack growth rate tests of candidate neutron-irradiated alloys in supercritical water. The tasks are described in the following sections.The work plan for this three year (FY05-FY07) program consists of four principal tasks; 1) the completion of a facility to conduct crack growth rate and constant extension rate tensile tests on highly radioactive, neutron irradiated samples in supercritical water, 2) constant extension rate tests and crack growth rate tests of candidate alloys in supercritical water, 3) proton irradiation and constant extension rate tests of proton-irradiated samples in supercritical water and 4) constant extension rate tests and crack growth rate tests of candidate neutron-irradiated alloys in supercritical water. The tasks are described in the following sections.
Highlights (Click an image for more information)
  • Material Selection for Supercritical Water Reactor

    The supercritical water reactor (SCWR) is one of the most promising Generation IV nuclear reactor concepts. Intergranular stress corrosion cracking (IGSCC) has been shown to be a major problem in current light water reactors, and so the higher operating temperatures and pressures anticipated for the SCWR creates a very challenging materials problem for this new reactor concept. In addition to the corrosive SCW coolant, the reactor component materials become damaged by neutron irradiation during reactor operation which causes irradiation assisted stress corrosion cracking (IASCC) which is one type of IGSCC. The IGSCC susceptibility of three candidate alloys selected for the SCWR concept (316L, 690, and 800H) were tested in study. The samples were irradiated with 3 MeV protons to emulate the effects of neutron irradiation damage and were then strained to failure in dearated SCW at a strain rate of 3×10-7 /s. The measurement used to quantify the susceptibility of the materials to IGSCC was the crack length per unit area on the gage surface of the specimens. The results of this test showed that austenitic stainless steel 316L was most susceptible to IGSCC in both the irradiated and unirradiated conditions, but the irradiation damage caused only a slight increase in the cracking susceptibility of the material. Nickel-base alloy 690 was shown to be the most resistant to IGSCC, but the crack length per unit area on the irradiated material was much higher than that on the unirradiated material indicating that the materials microstructure was not resistant to irradiation damage.