The current US fleet of nuclear reactors has been in service for three decades. Over this period, existing welds in stainless steel (SS) shrouds have sustained stress corrosion cracking (SCC) and are in need of repair. Additionally, helium has formed interstitially as a byproduct of proton bombardment. Current repair technology, such as TIG welding, puts extreme amounts of heat into the material and allows for interstitial helium atoms to aggregate and form bubbles/voids at grain boundaries. This significantly weakens the material, proving to be a very counterproductive and ineffective repair technique. Much study has been done on friction stir processing (FSP), but none has explored it as an enabling repair technology for use in nuclear applications. Because of its relatively low energy input as a solid state joining technology, it is proposed that FSP could effectively heal SCCs in these existing welds without the negative side effect of helium bubble formation. A spread of speeds and feeds were initially tested using a PCBN-W-Re tool on 304L SS. Six of these parameter sets were selected as representations of high, medium, and low temperature-per-power outputs for this research: 2 IPM-80 RPM, 2 IPM-150 RPM, 4 IPM-150 RPM, 4 IPM-250 RPM, 6 IPM-125 RPM, and 6 IPM-175 RPM. These varied parameter sets were tested for their tensile, micro-hardness, and corrosion resistant properties. In general, the lower IPM and RPM values resulted in higher ultimate tensile strengths (UTS). Higher IPM and RPM values resulted in tunnel, pin hole, and surface void defects. These defects caused premature failure in tensile tests and could often be identified through microscopy. Micro-hardness testing demonstrated a strong correlation per the Hall-Petch relationship – finer grain sizes resulted in higher yield strength (hardness values) of the material. The tool temperature during FSP was a good indicator of the expected hardness – lower temperatures resulted in higher hardness values. Corrosion testing was performed with a 1000-hour alternate immersion test in a room temperature 3.5% NaCl solution. With these testing parameters, the results demonstrated that FSP had no effect on the corrosion resistance of 304L SS under these conditions.

One of prominent issues related to failures in nuclear power components is attributed to material degradation due the aggressive environment conditions, and mechanical stresses. For instance, reactor core support components, such as fuel claddings, are under prolonged exposure to an intense neutron field from the fission of fuel and operate at elevated temperature under fatigue loadings caused by start up, shut down, and unscheduled emergency shut down. Additionally, exposure to highluence neutron radiation can lead to microscopic defects that result in material hardening and embrittlement, which significantly affects the physical and mechanical properties of the materials, resulting in further reduction in fatigue life of reactor structural components. The effects of fatigue damage on material deterioration can be further exacerbated by the presence of thermal loading, hold-time, and high-temperature water coolant environments. In this study, uniaxial fatigue models were used to predict fatigue behavior based only on simple monotonic properties including ultimate tensile strength and Brinell hardness. Two existing models, the Bäumel Seeger uniform material law and the Roessle Fatemi hardness method, were employed and extended to include the effects of test temperature, neutron irradiation fluence, irradiation induced helium and irradiation induced swellings on fatigue life of austenitic stainless steels. Furthermore, a methodology to estimate fatigue crack length using a strip-yield based model is presented. This methodology is also extended to address the effect of creep deformation in a presence of hold- times, and expanded to include the effects of irradiation and water environment. Reasonable fatigue life predictions and crack growth estimations are obtained for irradiated austenitic stainless steels types 304, 304L, and 316, when compared to the experimental data available in the literature. Lastly, a failure analysis methodology of a mixer unit shaft made of AISI 304 stainless steel is also presented using a conventional 14-step failure analysis approach. The primary mode of failure is identified to be intergranular stress cracking at the heat affected zones. A means of circumventing this type of failure in the future is presented.