In nuclear power plants, the corrosion products are generated from the surface of carbon steel piping and components. The corrosion products are transported into the secondary side of a stem generator (SG) from the feed water piping, and deposited on the surface of SG components such as SG tube, tube sheet, tube support plate. In order to remove the SG sludge, the sludge lancing and chemical cleaning should be performed periodically during the planned preventive maintenance. After the sludge lancing, the SG sludge samples could be collected and characterized using various analysis techniques. The SG sludge characteristic analysis results could be utilized to predict the integrity of SGs. The purpose of this study is to analyze the microstructure of the SG sludge collected from an operating pressurized water reactor. The flake deposits were polyhedral or spherical in shape and mostly composed of magnetite and contained small amounts of trevorite, jacobsite, and metallic copper particles. The various impurities such as sulfur, sodium, chloride, phosphrous, copper, and lead were detected in SG deposit samples. The concentration of impurities within the micro-pores of the flake deposits was evaluated to be 2 × 10⁶~3 × 10⁷ for copper and 2 × 10⁵ ~5 × 10⁵ for sulfur, sodium, and chloride. We calculated the impurity concentration factor within the micro-pores of SG deposits. When the metallic Cu and Pb particles and SG materials are electrically contacted and exposed the same electrolyte, Cu and Pb would behave as the cathode of the galvanic cell, while Alloy 600 and Alloy 690 would act as the anode and increase the corrosion rate. In addition, the stable states of impurities such as copper, lead, sulfur, and chloride were calculated in the secondary coolant condition of SG using thermodynamic calculation software. We also discussed the corrosion behavior prediction at the interface between the SG tube surface and the sludge.

Neutron Absorber Materials (NAMs) are mandatory for safety of spent nuclear fuel storage and transportation. Currently-used NAMs have some drawbacks such as requiring unnecessarily complex and uneconomical structural designs and highlighting the need to develop neutron absorbers with superior mechanical properties. Gadolinium (Gd) is notable for its high neutron absorption cross-section in the thermal neutron energy range, making it an excellent neutron absorber that can achieve effective neutron attenuation even in small quantities. In this study, titanium (Ti) was selected as the matrix metal for the development of Gd-containing NAMs. Four types of Ti-based NAM alloys were fabricated with 9 at.% Mo, 2 at.% Gd, and partial additions of Zr and Fe. Despite the presence of a small amount of the ω-phase, the Base alloy exhibited considerable tensile elongation and excellent work hardening behavior. Both Zr and Fe additions acted as β-Ti stabilizers, suppressing ω-phase formation. The tensile curves also reflected the influence of Zr or Fe additions, resulting in increased yield strength and tensile strength but reduced work hardening capability. This phenomenon is interpreted as a transition in the dominant deformation mechanism from twinning to dislocation slip.

Electron Beam Welding (EBW) is emerging as an innovative manufacturing technology to enhance the fabrication efficiency of Small Modular Reactor's (SMR) Reactor Pressure Vessels (RPV). The high energy density and rapid cooling characteristic of EBW necessitates the development of accurate prediction techniques for the resulting weldment microstructure and properties. This study simulated the EBW process of SA508 Gr.3 low-alloy steel using Finite Element Analysis (FEA) to predict the post-weld microstructure and hardness. Thermal analysis was conducted with ABAQUS software, and microstructure prediction was performed via user subroutines employing phase transformation and grain growth kinetic models. The thermal analysis results exhibited similarity to the the actual weldment geometry. Microstructure predictions anticipated martensite predominantly near the fusion zone (FZ) and bainite in the heat-affected zone (HAZ). However, this differed from experimental observations (which showed some martensite in the coarse grained HAZ among predominantly bainite and Widmanstantten ferrite), specifically regarding the primary location of martensite formation. The predicted hardness also showed a significant increase within the weldment, but the peak hardness value and location deviated from experimental results. Analysis of the simulation parameters highlighted the importance of accurately modeling the cooling rate and austenite grain size for prediction accuracy. Upon adjusting cooling conditions, the overall martensite fraction was significantly reduced, favoring bainite formation. Crucially, this led to martensite formation appearing relatively more within the HAZ region, aligning more closely with experimental observations. Consequently, the predicted hardness values also became comparable to experimental findings. The EBW model and prediction methodology developed in this study can serve as a fundamental technology for the analysis and optimization of weldments in SA508 low-alloy steel.

This study presents quantitative comparison of plasticity correction methods for environmental fatigue life assessment in primary stress-dominated structures under equi-biaxial loading. Conventional ASME Code procedures apply a simplified plastic correction factor (K_e), which may lead to overly conservative fatigue life predictions. To improve accuracy, equivalent strain energy density-based correction methods, Neuber and Glinka, were applied and compared using finite element analysis results. Fatigue tests were conducted with FABIME2e disk specimens to simulate equi-biaxial stress conditions, and the predicted fatigue lives were compared with experimental data. The results show that the ASME K_e method significantly underestimates fatigue life due to overpredicted plastic strain. In contrast, the Neuber and Glinka methods provided more realistic predictions and reduced conservatism. The selected plasticity correction method also affected the environmental fatigue correction factor (F_en), which depends on strain rate. This study confirms that equivalent strain energy density-based corrections are more suitable than conventional methods for evaluating environmental fatigue under equi-biaxial, primary stress-dominated conditions.

In Small Modular Reactors (SMRs), integral designs result in significant secondary stresses due to increased structural constraints under thermal loading. Conventional ASME Code-based fatigue evaluations, relying on elastic analysis and simplified plasticity correction factors, tend to overestimate fatigue damage under such conditions. Moreover, in cases where secondary stresses dominate, overestimation of fatigue damage becomes further amplified when environmental correction factors are applied, as prescribed in NRC Reg. Guide 1.207, Rev.1. This study investigates various plasticity correction methods, including simplified ASME approaches, Neuber method, Glinka method and elastic follow-up for their applicability to environmental fatigue evaluation of a thermally loaded stepped pipe specimen tested at Bettis. Finite element analysis under cyclic thermal loads was performed, and fatigue lives predicted using each method were compared against experimental crack initiation data. The results indicate that ASME Code-based procedures yield overly conservative predictions, while methods incorporating secondary stress redistribution provide fatigue life estimates that closely match experimental results. These findings support the adoption of advanced plasticity correction techniques to reduce conservatism in environmental fatigue evaluations for SMR structures.

This study conducted heating and cooling tests on a pipe-simulating specimen using a thermal fatigue testing system under various testing parameters. The effect of each parameter on the thermal fatigue test was investigated by comparing the temperature distributions across the specimen wall. The parameters tested were cooling time, the heating current of the induction heater, and the flow rate of coolant. The test results showed that the temperature distribution was nearly uniform when heating the specimen, and that it was less affected by the testing parameters. However, a significant temperature gradient was generated during cooling, which was affected by the testing parameters. In particular, cooling time had a considerably greater effect on the temperature distribution than the heating current and coolant flow rate.

SA508 Gr.3 Cl.1 steel is widely used as a material for reactor pressure vessels (RPVs) due to its high strength and excellent weldability. Electron Beam Welding (EBW) enables deep penetration with minimal heat distortion; however, unlike multi-pass arc welding, it lacks successive tempering effects, resulting in significant residual stress and microstructural inhomogeneity in the weld zone. Conventional Post-Weld Heat Treatment (PWHT) shows limited effectiveness in EBW joints and requires frequent non-destructive inspections during service. To address these limitations and improve structural reliability, this study proposes a Post-Weld Quality Heat Treatment (PWQHT) process as an alternative heat treatment method. The PWQHT process consists of two stages: austenitizing at 880℃ followed by tempering at 655℃. Experimental results showed that PWQHT promoted grain refinement and microstructural homogenization, leading to improved impact toughness and more uniform hardness distribution. In particular, the water quenching condition (WQ) resulted in the formation of tempered martensite, significantly enhancing toughness. In contrast, the air cooling condition (AC), which better reflects realistic industrial environments, produced tempered bainite with slightly lower mechanical properties but demonstrated greater practical applicability. This study provides valuable data demonstrating that the PWQHT process effectively improves the mechanical performance of EBW joints and contributes to ensuring the long-term operational reliability of critical structures such as reactor pressure vessels.

This paper explains procedures of design, manufacture, and performance verification of the nuclear power plant safety depressurization system (SDS) safety related valves. The existing check valve continuously caused seat leakage. So the pressure of the isolated pipe at the rear end increased to the operating pressure. The double sealing check valve developed through research has strengthened sealing power to prevent seat leakage. The developed double sealing check valve has been tested many times to operate at the designed differential pressure. The performance verification was carried out through the ASME QME-1 test on the motor actuated valve. Localization of safety depressurization system safety-related valves has been made to help the development of the nuclear industry.