Current phenomenological knowledge and understanding of mechanisms are reviewed for radiation embrittlement of reactor pressure vessel low alloy steels and irradiation assisted stress corrosion cracking of core internals of stainless steels. Accumulated test data of irradiated materials in light water reactors and microscopic analyses by using state-of-the-art techniques such as a three-dimensional atom probe and electron backscatter diffraction have significantly increased knowledge about microstructural features. Characteristics of solute clusters and deformation microstructures and their contributions to macroscopic material property changes have been clarified to a large extent, which provide keys to understand in the degradation mechanisms. However, there are still fundamental research issues that merit study for long-term operation of reactors that requires reliable quantitative prediction of radiation-induced degradation of component materials in low-dose rate high-dose conditions.

https://www.tandfonline.com/doi/pdf/10.1080/00223131.2013.772448

Current understanding of radiation-induced degradation in light water reactor structural materials

Evolution of manganese–nickel–silicon-dominated phases in highly irradiated reactor pressure vessel steels

Nuclear plant life extension to 80 years will require accurate predictions of neutron irradiation-induced increases in the ductile-brittle transition temperature (ΔT) of reactor pressure vessel steels at high fluence conditions that are far outside the existing database. Remarkable progress in mechanistic understanding of irradiation embrittlement has led to physically motivated ΔT correlation models that provide excellent statistical fits to the existing surveillance database. However, an important challenge is developing advanced embrittlement models for low flux-high fluence conditions pertinent to extended life. These new models must also provide better treatment of key variables and variable combinations and account for possible delayed formation of “late blooming phases” in low copper steels. Other issues include uncertainties in the compositions of actual vessel steels, methods to predict ΔT attenuation away from the reactor core, verification of the master curve method to directly measure the fracture toughness with small specimens and predicting ΔT for vessel annealing remediation and re-irradiation cycles.

Predictive reactor pressure vessel steel irradiation embrittlement models: Issues and opportunities

Atom probe tomography characterizations of high nickel, low copper surveillance RPV welds irradiated to high fluences

Nonlinear ultrasound was used to monitor radiation damage in two reactor pressure vessel (RPV) steels. The microstructural changes associated with radiation damage include changes in dislocation density and the formation of precipitates, and nonlinear ultrasonic waves are known to be sensitive to such changes. Six samples each of two different RPV steels were previously irradiated in the Rheinsberg power reactor to two fluence levels, up to 1020 n/cm2 (E > 1 MeV). Longitudinal waves were used to measure the acoustic nonlinearity in these samples, and the results show a clear increase in the measured acoustic nonlinearity from the unirradiated state to the medium dose, and then a decrease from medium dose to high dose.

Evaluation of radiation damage using nonlinear ultrasound

Improvement Of Radiation Embrittlement Dependences For VVER-1000 Pressure Vessel Materials On Lifetime Extension

Development of miniaturized specimens for the study of neutron irradiation/plasma exposure synergistic effects on candidate fusion reactor materials

Procedure of accelerated irradiation results usage for prediction of the material state corresponding to long-term operation of the reactor pressure vessel

After ten years of operation, the absorption tower at a nitric acid plant was failed due to stress corrosion cracking. SCC was caused by combination of both welding residual tensile stresses and chloride ions accumulated due to rain and/or vapor contaminated with chloride. Irregularity of fillet weld surface helped in occurrence of cycles of wetting and drying conditions of weld surface that in turn resulted in chloride concentration. The problem has been solved by changing design of the fixation system of the rain water collector in order to avoid combination of tensile stresses and chloride, the condition necessary for occurrence of SCC. In other words, fixation of rain water’ collector to the tower’ shell was done using mechanical means instead of fillet welds that in turn result in eliminating tensile stresses and reducing chloride concentration.

5639 - practical solution for avoiding failure of absorption tower at a nitric acid plant - replacing welding by mechanical joining -

Evolution of the Hydride Structure in Irradiated E110 Alloy in the Process of Thermomechanical Treatment Simulating Supercritical Dry Storage Conditions

O. O. Zabusov

Papers

Improvement Of Radiation Embrittlement Dependences For VVER-1000 Pressure Vessel Materials On Lifetime Extension

Development of miniaturized specimens for the study of neutron irradiation/plasma exposure synergistic effects on candidate fusion reactor materials

Procedure of accelerated irradiation results usage for prediction of the material state corresponding to long-term operation of the reactor pressure vessel

5639 - practical solution for avoiding failure of absorption tower at a nitric acid plant - replacing welding by mechanical joining -

Evolution of the Hydride Structure in Irradiated E110 Alloy in the Process of Thermomechanical Treatment Simulating Supercritical Dry Storage Conditions