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Example of stress corrosion cracking of a Type 316 stainless steel under thermal insulation at 50 to 60°C. Rinsing water containing 60 mg/kg of chloride and residual stresses are at the origin of cracking, [During, 1991]. 

Example of stress corrosion cracking of a Type 316 stainless steel under thermal insulation at 50 to 60°C. Rinsing water containing 60 mg/kg of chloride and residual stresses are at the origin of cracking, [During, 1991]. 

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... Zirconium alloys are readily weldable but as with all reactive metals need special precautions to prevent pickup of interstitial elements like oxygen, carbon, and nitrogen that can degrade both the mechanical properties and the corrosion performance of the weld. [110][111][112] Vacuum or inert gases (argon, helium, or Ar-He mixtures) can be used to shield zirconium but again, care needs to be taken to ensure sufficient vacuum level or gas purity to prevent contamination. 110,112 Zirconium alloys can be susceptible to both supersolidus and subsolidus (e.g., hydride-type) cracking. ...
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Chapter
Nuclear power systems are constructed from a wide range of metallic alloys, subjected to taxing environmental conditions, and required to resist cracking and degradation of their principal mechanical and physical properties for decades. Fusion welding is, in general, the joining method of choice due to its hermeticity, high joint efficiency, and economic advantages relative to mechanical or brazed joints. However, it is often fusion welds, or their heat affected zones that prematurely degrade or fail due to the complex interplay of physical defects, compositional gradients, metallurgical changes, and residual stresses. This chapter presents the current mechanistic understanding of welding defects, reviews recent developments in assessing residual stresses & plastic strains and relates these factors to the in-service performance of welds. Lastly, the weldability of common structural alloy systems is reviewed.
... As seen in the mass gain plots, breakaway oxidation occurs in all three alloys at temperatures ≥700°C and Zry-4 experiences breakaway the earliest of the three alloys, while Zr-1Nb is the most resistant to breakaway. From Figs. 5-7, and as expected from the discussion above, the HAZ is larger in the TIG welded tubes than in the EBW specimens [18]. The effect of plastic deformation on oxidation behavior is seen in the Sn-Fe containing alloys (Zry-3 and Zry-4); accelerated oxidation on the chamfer regions are clearly visible. ...
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ODS steels (Oxide Dispersion Strengthened) are candidate materials for fuel cladding in Sodium Fast Reactors (SFR), one of the concepts at study for the forth generation of nuclear power plant. These materials have good mechanical properties at high temperature due to a dispersion of nanometer-sized oxides into the matrix. Previous studies have shown that melting can induce a decrease of the mechanical properties at high temperatures due to modifications of the nanometer-sized oxide dispersion. Therefore the fusion welding techniques are not recommended and the solid state bonding has to be evaluated. This study is focused on resistance upset welding.Welding experiments and numerical simulations of the process are coupled in this thesis. All the trials (experimental and numerical) are built using the experimental design method in order to evaluate the effects of the process parameters on the welding and on the weld. A 20Cr ODS steel is used in order to conduct the study.The first part is dedicated to the study of the influence of the process parameters on the welding. The simulation shows that the welding steps can be divided in three stages. First, the temperature of the contact between pieces increases. Second, the process is driven by the pieces geometry and especially the current constriction due to the thinness of the clad compare to the massive plug. Therefore, the heat generation is mainly located in the clad part out of the electrode leading to its collapse which is the third stage of the welding step. The evaluation of the process parameters influences on the physical phenomena (thermal, mechanical ...) occurring during the welding step allow to adjust them in order to influence the thermal and mechanical solicitation undergone by the pieces during the welding process.The second part is dedicated to the study of the influence of the physical phenomena on the welds. In the process parameter range, some welds exhibit compactness defects or a modification of the microstructure and of the oxide dispersion. Compactness defects are related to thermal and mechanical phenomena occurring at the contact between pieces. The modification of the microstructure is related to dynamical recrystallization or to a local fusion. The dynamical recrystallization occurring in the clad due to high deformation and high temperature is linked to modification of the oxide dispersion.Using the effects of the process parameters on the welding and on the weld, it is possible to adjust the temperature and the deformation in order to avoid the compactness defects and the modification of the oxide dispersion. All these results are then apply to the welding of a 9Cr ODS steel which is a candidate alloy for the SFR fuel cladding. The effects of the material properties on the welding and the weld are then discussed by comparing the two alloy with a different chromium content but also by comparing results on the 20Cr-ODS with a material of similar chemical composition but without the oxide dispersion.
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Among the factors influencing the degradation of the spent nuclear fuel cladding in interim dry storage, the irradiation history and the average burnup at discharge must be considered. In fact, due to the pellet-cladding contact particularly affecting high burnup fuels, the inner surface of the cladding becomes increasingly exposed to the damage caused by the alpha decaying actinides present at the rim of the pellet. Moreover, due to the low temperature conditions characteristic of the interim dry storage, thermal recombination of the produced defects is not expected to occur. Here, we investigate the irradiation damage accumulated inside an irradiated Zircaloy-4 cladding 32, 55 and 100 years after the end of irradiation and discharge from the reactor core. The considered cladded UO2 pellet belongs to a Pressurised Water Reactor (PWR) fuel rod consisting of five segments and having an average burnup at discharge of 50.4 GWdtHM−1. The calculations performed with Fluka 2021.2.0 Monte Carlo code show that the volume mostly affected by the irradiation damage corresponds to the ZrO2 layer formed between the pellet and the cladding. The actinides which are responsible for the alpha damage are mainly 242Cm, 244Cm, 241Pu and 238Pu. The recoiling daughter nuclei during the alpha decays produce irradiation damage only within the first μm of oxide layer.
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In the present study, Zr-2.5 wt.%Nb alloy plates of thickness 4 mm have been welded using pulsed Nd:YAG laser system at different process parameters and characterized in terms of microstructural evolution, hardness, residual stress, and mechanical properties. It has been observed that heat input plays an important role on mode of welding. The full penetration up to a thickness of 4 mm of the alloy has been achieved at a minimum laser heat input of 800 J mm⁻¹ without any crack and porosity formation. For avoiding the porosity formation in the weld zones, a transition mode between conduction and keyhole has been used by optimizing laser process parameters. The microstructural analysis revealed that the fusion (FZ) is consists of predominant lath type α′ martensitic phase with small amount of randomly distributed acicular type of α′ martensite phase. However, the heat affected zone (HAZ) have lath type α′ martensitic phase together with αZr phase. Further, in the FZ and HAZ regions, the presence of retained βZr phase is higher as compared to the base metal (BM). The change in microstructure and phase field of different weld zones has been explained by evaluating the time-temperature profiles and cooling rates using COMSOL multi-physics simulation. In addition, the FZ and HAZ zones have been found to have tensile residual stress of the order of 280 MPa and 145 MPa as compared to the BM (-70 MPa). The microhardness in the FZ region has been observed to be higher (240-260 HV0.1) as compared to the BM (∼185 HV0.1) due to the formation of martensitic phase. The tensile room temperature testing showed that the mechanical strength of as welded sample is significantly higher than the base metal with lower ductility. The fractography of the fractured surfaces confirmed ductile nature of failure in the as welded samples.