Irradiation-Assisted Stress Corrosion Cracking
Irradiation-assisted stress corrosion cracking (IASCC) is one type of materials degradation of particular interest to our group. IASCC refers to the premature failure of light water reactor (LWR) core components under the combined action of applied stress, a corrosive environment and irradiation. This form of material degradation has been occurring for over 30 years in reactor core internal components made of both austenitic iron and nickel-base alloys. It is becoming increasingly evident that the problem is widespread and that many more components are susceptible to this form of degradation.
Irradiation-assisted stress corrosion cracking poses a significant economic penalty for nuclear power reactors. Although most failed components can be repaired or replaced, such operations are difficult and expensive. Worse, IASCC susceptibility increases with increasing radiation dose. So, in addition to the components that already exhibit cracking, many more components may become susceptible in the near future. This makes IASCC a significant problem when determining whether a given nuclear power plant can be operated to the initial design life, much less an extended lifetime.
Understanding the mechanisms that control IASCC is an important step in designing materials that are resistant to this form of degradation and for developing mitigation strategies for existing materials and reactors. Radiation induces a number of changes in core components (most notoriously, microchemical effects, microstructural changes, and hardening) which all may influence cracking susceptibility. However, despite years of research, the underlying mechanisms of IASCC are still not understood.
Stress Corrosion Cracking in Supercritical Water
The supercritical water reactor (SCWR) is one of the Generation IV nuclear reactor concepts. The increased efficiency and simplified design of the reactor make the concept promising, but the reactor environment will be very severe. Temperatures are expected to be as high as 620ºC and internal core components will be exposed to doses as high as 15 dpa. The high temperature aqueous environment and irradiation damage to metallic structural components creates the potential for high corrosion rates and irradiation assisted stress corrosion cracking (IASCC).
Scanning Electron Microscope Images of 316L, D9, and 690
The corrosion and IASCC behaviors of several of the candidate alloys for the reactor concept have been evaluated in the High Temperature Corrosion Laboratory. The alloy classes tested include zirconium alloys, austenitic stainless steels, Ni-base alloys, ferritic-martensitic steels, and oxide dispersion strengthened alloys. Studies on zirconium alloys Zircaloy-2 and Zircaloy-4 showed extreme oxidation and indicated the alloys were not viable for the SCWR. The large majority of the ferritic-martensitic steels were not susceptible to IASCC, but their oxidation behavior may limit their use for the SCWR. The austenitic stainless steels and Ni-base alloys were resistant to oxidation, but all have shown susceptibility to intergranular stress corrosion cracking (IGSCC) that intensified with irradiation damage. Materials issues are a major challenge for the SCWR and the High Temperature Corrosion Laboratory is one of only a few laboratories worldwide with the capability to perform the corrosion and SCC experiments.
Materials for Extreme Environments in Advanced Nuclear Reactor Systems
The combination of high temperature, chemically aggressive coolants, a high radiation flux and mechanical stress poses a major challenge for the development of advanced fission/ fusion reactor concepts. In particular, the confluence of these four drivers presents a unique and extremely challenging environment for materials. The greatest challenges to extending the life of the current commercial fleet of reactors are corrosion and stress corrosion cracking of core components exposed to high temperature water and high radiation flux. It is these same degradation processes that are crucial to the success of advanced reactor concepts for the 21st century. However, we don’t understand the underpinning degradation processes in the complex environment consisting of radiation, a corrosive coolant and mechanical stress at high temperature.
Scanning Electron Microscope Images of 304, 316L, 625, and 690 @ 550°C
This deficiency in understanding is caused by the complicated synergisms among environmental variables and an historical lack of attention to fundamental environmental degradation processes in reactor system environments. Knowing how radiation influences the basic mechanisms of film formation, growth, breakdown and repair along with species transport is key to understanding materials compatibility issues in all advanced reactor concepts, from which new, resistant materials can be developed. The challenge then, is to understand how materials degrade in complicated, extreme environments characterized by high temperature, aggressive media, radiation and applied stress, and then to use that understanding to create materials and surfaces that are stable and non-reactive in these same environments.