Papers by Author: Wei Wei Yu

Abstract: Most of the French Nuclear Power Plants (NPPs) are currently embarking upon efforts to renew their operating license, while the pressurized thermal shock (PTS) events and environmentally assisted fatigue (EAF) pose potentially significant challenges to the structural integrity of the reactor pressure vessel (RPV) which has the potential to be NPP life-limiting conditions. In the assessment of the PTS events, the deterministic fracture mechanics (DFM) is still used as the basic mechanics in most countries except for the USA. While the maximum nil-ductility-transition temperature (RT_NDT) is about 80°C for 54 French RPVs after 40 years operation, the maximum allowable RT_NDT is only about 70 ^oC and 80 ^oC for the typical PTS events in the IAEA and NEA reports, respectively. On the other hand, the effects of light water reactor (LWR) environmental (other than moderate environment in the code) were not considered in the original design, while the effects of LWR environmental are needed to be considered in the LRA according to the USA regulations. In this paper, the challenges of the PTS and EAF are discussed, and some suggestions are also given for the LRA

Abstract: In order to simulate the thermal influence on springback property of O-ring, based on elastic-plastic contact analysis method of finite element analysis (FEA), numerical simulation of compression-springback course of O-ring was carried out by ANSYS software. The results show that it causes a sharp decrease in the peak force applied on O-ring, corresponding to a decline on the contact stress distribution which may influence the seal ability of the O-ring. The simulation results are in good agreement with the experimental results.

Abstract: In this paper, a series of bending cyclic tests under stress controlled were conducted at room temperature on Zircaloy-4 (Zr-4) to investigate its bending ratcheting behavior. The effects of mean stress and stress amplitude on the bending ratcheting behavior were experimentally studied, respectively. The experimental results show that the ratcheting strain of the material is very sensitive to mean stress and stress amplitude. It can be concluded that ratcheting strain level increases with increasing mean stress and stress amplitude.

Abstract: Hoop creep tests and axial creep tests were carried out on a new zirconium alloy for nuclear fuel cladding tubes at 375 °C. The result indicated that the proportion of the second creep stage decreased as the stress increasing, meanwhile, the steady-state creep rate increased in both creep test modes. However, the ability of creep rupture resistance in the hoop direction is more strengthen than that in axial direction. The SEM analysis showed that the creep fracture was mainly dimple toughness fracture. The dimple for low stress was intensive and small. The fitting curve by the θ function was well coincided with the test data, so the creep life can be analyzed by θ function.

Abstract: To investigate the property of a new type of Zircaloy material, a low cycle fatigue (LCF) test has been performed at room temperature (RT) and 375°C. Results show that the new alloy generally displays cyclic hardening followed by a continuous softening behavior. Fatigue lifetime curves as a function of strain range imply that the new alloy has a nearly same lifetime than that of Zr-4 at RT, and superior than that at 375°C.

Abstract: In this paper, the role of the pearlite-banded structure on fatigue crack growth behavior was investigated on carbon vessel plate material SA516, which is commonly used in the nuclear power plants. Along pearlite-banded orientation, in situ fatigue tests indicate that the crack initiated and propagated in the ferrite and then extended along the ferrite-pearlite interface when it met pearlitic colony. For comparison, the cyclic loading was also carried out perpendicular to the banding direction of the microstructure, and an intense crack branching was observed which led to fatigue crack retardation. Besides, the orientation perpendicular to banded pearlite in the investigated ferrite-pearlite steel was found to have a lower fatigue crack growth rate.

Abstract: In order to confirm the structural integrity of pressurizer surge line affected by thermal stratification and thermal shock, the thermo-mechanical fatigue (TMF) behavior of the material used for surge line was investigated based on the real situation in the pressurized water reactor (PWR). Smooth, hollow specimens were subjected to in-phase (IP) and out-of-phase (OP) cycling in air under a mechanical strain control mode. For the sake of comparison, low cycle fatigue (LCF) tests were also performed at the peak temperatures of TMF cycling. The Nano Hardness Tester was used to test the nano hardness of the sample on the cut section surface. The results are shown that there is no significant difference between the IP, OP and IF lives in the investigated temperature ranges. The fracture analysis reveals that the crack initiation and propagation occurred in a transgranular mode under OP, IP and IF cycling condition, and a harden layer occurrence may be the cause of the crack initiation.

Abstract: In this study, based on ASME Code Section III, France RCC-MR, ANSYS simulation and experimental data, ratcheting and ratcheting boundary are obtained to determine allowed loading of primary auxiliary pipeline in nuclear power plant. A software of pressurized water reactor auxiliary pipeline anti-ratcheting design is compiled. The calculation method to determine the boundary of ratcheting effect is proposed in order to provide a basis for pipe structural design. This work focuses on the work of the establishment of the structure of software system. The structure mainly composed by three program modules, i.e. the database module, data computing module and data display module. The function of each module and the relationship between the modules are explained. An example to determine ratcheting boundary is presented in the paper.

Abstract: A series of ratcheting experiments under bending loading for Z2CND18.12 stainless steel were carried out on a dynamic mechanical analyzer. The characteristic of progressive inelastic deformation of the material in the different temperatures and different bending loadings were investigated. The difference of the domestic and imported stainless steel on the bending ratcheting properties was studied. It is shown that the temperature and the bending loading affect the ratcheting greatly. In addition, the imported steel is better than the domestic one on the resistance to the ratcheting deformation at various temperatures and bending loadings.

Abstract: In order to determine the ductile to brittle transition behavior of nuclear primary pipe material (Z3CN20.09M) during the thermal aging procedure, instrumented impact tests at different temperatures were performed on Z3CN20.09M aged at 400°C for up to 3000 hours. The load-deflection curves from the instrumented impact tests described both the loading and the fracture stages, from which the dynamic strengths and energy were calculated. The results indicated that the thermal aging decreases the impact energy and shifts the ductile-to-brittle transition curves to higher temperatures.