Mechanism and Kinetics of Thermal Embrittlement of Austenitic Chromium-Nickel Steel during Long-Term High-Temperature Holding

This article describes a model of structural changes in austenitic chromium-nickel steel during long-term thermal soakings. The time for reaching equilibrium content of Cr23C6 carbide at a temperature of 600 °C is calculated. The calculation results are compared with experimental data on the kinetics of thermal embrittlement of steel at a temperature of 600 °C after thermal holding for up to 15000 hours.

Flaw Evaluation Procedure for Cast Austenitic Stainless Steel Materials Using a Newly Developed Statistical Thermal Aging Model

Thermal embrittlement of some cast austenitic stainless steels (CASS) occurs at reactor operating temperatures can lead to a reduction in the fracture toughness and increase in strength. Some aged CASS materials have the potential to have exceedingly low toughness and also show high variability due to the nature of their microstructure or compositional variation within the casting. Because of their low aged toughness with the variability, flaw evaluations of CASS material need to be done with an understanding of the materials aged condition, especially since most US PWR nuclear plants have been given plant life extensions for 60-year operation, and consideration of further extension to 80 years is underway. In this paper, a flaw evaluation procedure for CASS materials is presented using a new statistical model developed to predict the toughness of fully aged CASS using the material’s chemical composition. The new statistical model was developed based on the experimental toughness using standard 1T CT specimens (generally in the L-C orientation) at 288C to 320C and chemical compositions of the CF8m CASS materials. While the detail development of the model is beyond the scope of this paper, a brief validation of predicted toughness using chemical compositions is presented in this paper. Using the predicted toughness, a flaw evaluation procedure was developed using the Dimensionless-Plastic-Zone-Parameter (DPZP) analysis to determine when limit-load is applicable and also approximate the elastic-plastic correction factor (Z-factor) that needs to be applied to the limit-load solution to predict the failure stress for CASS pipe and fittings with a circumferential surface crack. Variability within a single casting was also determined from available test results which was included in the procedure to determine Z-factor. The procedure was then validated against several CF8m pipe test results which include various pipe diameters, crack sizes, ferrite contents, failure modes (i.e., limit load or EPFM), etc. The as-developed flaw evaluation procedure was also used to determine the Z-factors for four different pipe diameters for a database of 274 pipe/elbows in US PWR plants (whose chemical compositions were known) — essentially solving 1096 sample problems to understand what range of Z-factors might exists in US PWR plants (for CF8m CASS materials) considering all variations in pipe dimensions, ferrite contents, materials’ toughness, etc. Finally, the applicability of the CF8m-based statistical model for use with CF3 and CF8 CASS materials was also investigated by comparing the predictions with available test results.

Consideration of Thermal Embrittlement in Alloy 316H for Advanced Non-Light Water Reactor Applications

To construct advanced non-light water reactors (ANLWRs) operating in the temperature range above that for the traditional light water reactors (LWRs), Alloy 316H is one of the candidate materials because of its inexpensiveness, significant service experience, and qualification for nuclear applications by the American Society of Mechanical Engineers (ASME). However, during the life span at temperatures expected for the ANLWRs, the alloy is likely to experience thermal embrittlement that has not been a concern for the traditional LWRs. To prepare for the development, the possibility of adverse thermal embrittlement effects on Alloy 316H performance in the ANLWRs must be evaluated and a technical basis regarding thermal embrittlement, if necessary, must be established for structural integrity analysis to provide reasonable assurance of adequate nuclear safety protection. In this paper, current technical basis for nuclear applications of Alloy 316H deterioration from thermal aging is briefly introduced. The likelihood of adverse thermal embrittlement effects on Alloy 316H performance is evaluated through historical data on microstructural and mechanical property evolution. Characterization of thermal embrittlement is then discussed, followed by a review of predictive models and trend curves for alloy embrittlement. Based on the review and evaluation, technical gaps for addressing thermal embrittlement issues are identified and gap-filling actions are recommended for establishing a technical basis to enable adequate consideration of thermal embrittlement in Alloy 316H applications to the ANLWRs.

Evolution of Impact Properties of 16MND5 Forgings for Nuclear Reactor Pressure Vessel during Thermal Aging at 500°C

Effects of thermal aging on tensile and Charpy impact properties in 16MND5 steel was investigated, which were aged at 500°C for 0 h, 1000 h, 3000 h, 5000 h. A significant decrease in the yield stress and ultimate tensile strength was observed after thermal aging, while the elongation exhibited a slight decrease follow by an increase aged for 5000 h. What's more, the ductile-to-brittle transition temperature (DBTT) showed a remarkable increase with the prolongation of thermal aging duration. These facts indicate thermal aging caused embrittlement of the steel, which was further investigated by microstructure observation of SEM. The results show cleavage fracture after thermal aging. Furthermore, experimental results at 350°C thermal aging temperature originated from the previous literature were used to analysis the effect of thermal aging temperature. Thus, thermal embrittlement should be taken seriously.