In this paper, a method to quantitatively predict the reduction in fracture resistance with thermal embrittlement using micro indentation test. The change in hardness of δ-ferrite with thermal embrittlement is characterized using a concept of hardening constant. Then, thermal embrittlement constant to predict the reduction in fracture resistance with thermal embrittlement is derived from the relation between hardening constant and thermal embrittlement constant. Finally, the reduction in fracture resistance with thermal embrittlement is predicted by multiplying thermal embrittlement constant obtained from micro indentation test by J-R curve of unaged material. The validity of the hardening constant concept is confirmed by comparing the prediction and experimental results.

This study proposes a practical adjustment equation for the Embrittlement Trend Curve (ETC) to effectively apply it to reactor pressure vessel (RPV) materials in individual nuclear power plants. Traditional ETC adjustment methods have limitations due to constraints in number of group-specific measurements and a lack of statistical foundations. To address these issues, KAERI applied a Markov Chain Monte Carlo (MCMC)-based mixed-effect model to the latest ETC model, ASTM E900-15. This approach quantitatively calculates the mean, standard deviation, and prediction intervals of the adjustment intercept by considering the grouping characteristics of surveillance data and uncertainties in unirradiated specimens. Although the KAERI model provides quantitative distributions of parameters and intercepts, it has challenges in practical applications due to computational complexity and low portability. In this study, a simplified equation was developed using the statistical calculations of the mixed-effect model, which retains the primary outcomes of the KAERI model while enhancing portability. This equation supports effective adjustments to the ASTM E900-15 ETC for nuclear power plants with diverse material properties and operational conditions. It enables reliable evaluations of RPV integrity using plant-specific surveillance data. The findings of this study are expected to improve the precision and practicality of ETCs.

This study analyzes the effect of solid-state phase transformation (SSPT) on finite element (FE) residual stress analysis. The SA508 Grade 3 Class 1 plate was used for the analysis. A cylindrical 3D heat source was applied for the thermal simulation, and the thermal model variables were determined using experimental temperature variation over time and the shape of fusion zone after welding. Stress analysis was performed based on the temperature gradient obtained from the thermal simulation, and the experimental residual stress of electron beam-welded SA508 was accurately reproduced with consideration of SSPT during simulation. The validated model was then used to investigate the effect of SSPT on welding residual stress. The results showed that longitudinal residual stress at the weld centerline was significantly overestimated, however, the transverse residual stress was acceptably reproduced. Consequently, the transverse residual stress might be estimated without consideration of SSPT, and the longitudinal stress should be analyzed with consideration of SSPT to improve the design margin for electron beam-welded structures.

In this study, the influencing factors of the four simplified elastic-plastic analysis methods (ASME B&PV Code Sec. III NB, ASME Code Cases N-779, N-904 and JSME EPD Code Case) for thermal fatigue loading were investigated via FE analysis of a simulated specimen. As expected, the ASME B&PV Code Sec. III NB method is the most conservative, because it applies the same K_e-factor to both primary and secondary stresses. The JSME EPD Code Case predicts a fatigue life 1 to 2 times longer than the ASME Code Sec. III method, depending on the restraint condition. The Code Case N-904 predicts a fatigue life 1 to 5 times longer. Finally, the Code Case N-779 provides the greatest reduction in conservatism, with a predicted fatigue life 3 to 7 times longer than the ASME Code Sec. III method.

The International Atomic Energy Agency (IAEA) highlights the importance of maintaining the structural integrity of key safety equipments and facilities not only under design-basis earthquake but also beyond-design-basis earthquake. To ensure this integrity, it is essential to utilize appropriate elastic-plastic fatigue analysis methods for earthquake resistance. In this study, a parametric study of FE simulation for piping elbow under dynamic cyclic loading was conducted by varying the damping ratio and element density for the pipe elbow specimen for application of the JSME Code Case for elastic-plastic fatigue analysis. The elbow used in this research was made from SA403 WP316 stainless steel, used in primary and secondary piping systems of nuclear power plants. To simulate the pipe elbow test, 3-D elastic-plastic FE analysis were performed using the bi-linear kinematic hardening model. Dynamic cyclic loadings were applied at room temperature. The FE analysis was compared with experimental acceleration response and the opening-closing displacement data.

This study develops a thermal fatigue test system to simulate the fatigue damage in the pipes and nozzles of nuclear power plants with the pre-test to confirm its performance. The test system was designed to apply cyclic thermal loads by periodically injecting coolant into the interior of the cylindrical specimen with the outer surface continuously heated. The confirmation test shows that it can be applied not only to pure thermal loads but also to combined thermal and mechanical loads to the specimen. The system also has the heating and cooling capacity to control performance to cool and reheat to a specific temperature within a limited time. It is also confirmed that the system can successfully generate a large temperature gradient in the wall of the specimen, resulting in fatigue damage.

This study investigates the application of the extended finite element method (xFEM) for fatigue and stress corrosion crack growth analysis in nuclear piping. Stress corrosion crack growth analysis was performed using xFEM considering various element size ratio. The results are compared with ASME Section XI procedures and the AI-FEM program. An appropriate element size ratio was presented through the comparison of crack growth results. The proposed element size ratio was validated in the context of piping fatigue crack growth. The xFEM results applying the proposed element size criteria showed good agreement with ASME Section XI and AI-FEM for both stress corrosion and fatigue crack growth. Additionally, the applicability of xFEM to complex crack shapes is demonstrated through comparison with AI-FEM results. The study concludes that xFEM can effectively simulate crack growth in nuclear piping with various thicknesses, crack geometries, and loading conditions when using the proposed modeling criterion. This research contributes to enhancing the applicability of xFEM in flaw evaluations for nuclear piping, providing an alternative to traditional finite element methods and code-based procedures.

A computer code that can calculate pH and conductivity in high-temperature water chemistry was developed. In the code, chemical species used in simulated reactor water conditions were considered. With given ionic concentrations, pH and conductivity of water could be computed at a given temperature. The corrosion potential of austenitic stainless steels (SSs) was also estimated based on the mixed-potential theory. With the given pH and dissolved gas (hydrogen, oxygen, and hydrogen peroxide), the corrosion potential could be estimated at a given temperature. The validation results showed that the code yielded reasonably accurate predictions compared to the experimental data. Therefore, predicted pH, conductivity, and corrosion potential can be used as input features for predicting stress corrosion cracking growth rates of SSs in nuclear power plant conditions.

During long-term operation of pressurized water reactors (PWRs), hydrogen diffusion from the primary to the secondary side of the steam generator (SG) tube can occur due to the difference in hydrogen concentration between the primary coolant and the secondary coolant. Therefore, the secondary side of SGs in PWRs could be subject to this potential degradation mechanism. In particular, in the crevices between SG tube and tube support plate, the aggressive chemical impurities can be concentrated in the porous magnetite deposits to form the highly corrosive environment by means of local boiling. Despite the replacement of Alloy 600 tubes with Alloy 690 tubes, the general corrosion of SG tubes remains a potential problem that can reduce the integrity of SGs. The purpose of this investigation is to elucidate the general corrosion behavior of Alloy 600 and Alloy 690 SG tube in alkaline solution containing the impurities and hydrogen. The representative impurities such as Pb, Na, Cl and S were selected to simulate the chemistry of SG crevices. In both SG tube specimens, the polyhedral oxide particles were mainly observed with some round oxide particles. Compared with Alloy 600 tube specimen, the general corrosion rate of Alloy 690 significantly increased. In the case of the corrosion test for 3000 h, the corrosion rates of Alloy 690 tube specimen increased by about 7.8 times, compared with that of Alloy 600 tube specimen. The concentration of Pb contents of the oxide layers formed on Alloy 690 tube specimen was higher than that of Alloy 600 tube specimen. The mechanism of the difference in the general corrosion behavior between Alloy 600 and Alloy 690 SG tubes was discussed from the viewpoint of the thermodynamically consideration between alloying elements and Pb and E-pH diagram.

Transportable Micro Reactors (TMRs) harness the advantages of Small Modular Reactors (SMRs) while addressing the need for mobility and transient power supply. These factory-manufactured reactors can be quickly installed and relocated, enhancing their practicality. This study evaluates radiation safety during transportation, focusing on exposure from radioactive sources within the reactor vessel. While significant radiation from fission products in the active core can be effectively shielded by the inner reflector, the induced radiation from surrounding components requires careful assessment to ensure compliance with exposure limits for operators. Given the constraints on the size and weight of the transport container, minimizing fast neutron leakage is crucial. Utilizing OpenMC for depletion calculations, we identify six key isotopes-Cr-51, Fe-59, Mn-54, Mn-56, Co-58, and Co-60-generated during a conservative operational scenario. Our findings indicate that maintaining surface doses below 10 mSv/h requires fast neutron leakage from the inner surface of the reactor vessel to be under 1.0E+8/cm²·sec. Therefore, optimizing core size and reflector thickness is essential for the effective design of TMRs without the need for additional shielding.

The study proposes an improved application of the the Riera method, which based on trial and error method, to aircraft engine impact analysis on spent nuclear fuel transport casks. The improved application was applied to the case where an aircraft engine simulation model collides with a rigid wall, and finite element explicit dynamic analysis was performed. In addition, the results of the improved application method were compared with those of the full model and the existing Riera method. The comparison results confirmed that the proposed improvement application is an efficient and reliable method that matches well with the full model.

This paper evaluates the ultimate pressure capacity (UPC) of a prestressed concrete containment building, considering the loss of prestress. A three-dimensional finite element model with openings was utilized for the UPC evaluation, and material properties reflecting severe accident temperatures were applied. The failure criteria were established based on the strain of the liner plate in accordance with the regulatory guidelines. When the prestress was reduced by approximately 14% from its initial value, the UPC decreased by about 4%. This study acknowledges limitations in the determination of nonlinear material properties, realistic prestress losses, and failure criteria. Further research to address these limitations could enhance the accuracy of UPC evaluation.