Environmentally Assisted Cracking (EAC) of Reactor Materials

We perform experimental testing and modeling to assess the crack growth rates for structural materials in simulated light-water reactor environments. Four test facilities are being used to develop experimental data on primary water stress corrosion cracking (PWSCC) and corrosion fatigue behavior of nickel-alloys and their welds. Our results provide valuable information to further the understanding of stress corrosion cracking and provide data needed for reactor safety evaluations and regulations.

Additional Information »
NE Departments involved:
• Engineering Development and Applications

Corrosion Performance/Metal Dusting

We test structural materials for their corrosion resistance in simulated metal dusting environments at elevated temperatures, typical of conditions in hydrogen-, ammonia-, and methanol-reformers and in syngas plants. We also use atmospheric test facilities to develop a fundamental understanding of metal dusting reactions and material performance over a wide range of simulated process conditions, temperatures, surface finish and coatings. These facilities are used to test developmental alloys and coatings with improved metal dusting resistance.

Additional Information »
NE Departments involved:
• Engineering Development and Applications

Irradiated Materials

Two hot-cell test facilities are used to develop experimental data on the irradiation-assisted stress-corrosion cracking susceptibility, fracture toughness, and corrosion fatigue, of wrought and cast austenitic stainless steels and their welds. These results provide valuable information needed for reactor safety evaluations and regulations. Other hot cells are used to fabricate guide-tube and fuel-cladding mechanical properties test specimens and to conduct high-temperature steam oxidation and water quench tests using high-burnup cladding samples for Loss-of-Coolant Accident (LOCA) research.

Additional Information »
NE Departments involved:
• Engineering Development and Applications

Steam Generator Tube Integrity

We conduct research and development and validate models for tube performance in pressurized water reactor steam generators.

Steam Generator Tube Integrity Program
NE Departments involved:
• Engineering Development and Applications

  Other NE Capabilities:

• Nuclear Systems Modeling and Design Analysis
• Nuclear Systems Technologies
• Risk and Safety Assessments
• National Security & Non-Proliferation Technologies
• Materials Testing
• Engineering Computation & Design
• Engineering Experimentation