[en] Based on the design loads of the pressurizer in nuclear power plants, this paper investigates the crack propagation induced by the fatigue and the stress corrosion of the design overlay weld structure in the pressurizer nozzle safety end. The crack propagation depth is obtained, and the safety assessment is carried out. At the end of the cycle, the maximum circumferential crack growth depth and axial crack growth depth in dissimilar metal weld zone are 0.4 × 10^-3 mm and 23.6 × 10^-3 mm, respectively, while the maximum circumferential crack growth depth and axial crack growth depth in stainless steel weld zone are 12.4 × 10^-3 mm and 0 mm. The results show that the crack propagation meets the design requirement. This analysis can provide a reference for the overlay weld structure design and evaluation of the pressurizer nozzle safety end. (authors)

[en] During the core meltdown severe accident in the nuclear power plants, the strategy of In-Vessel Retention (IVR) to contain the melt in the Reactor Pressure Vessel (RPV) is a key mitigation measure. During the implementation of IVR strategy, the RPV lower head is likely to fail due to excessive creep deformation under the combined action of extremely high temperature loads and mechanical loads. Therefore, it is necessary to perform the analysis of the creep deformation of the RPV lower head, to ensure the structural integrity of the RPV under the condition of melt retention. In this paper, under the assumption of IVR, the finite element method is used to perform the thermal-structural coupling analysis of the RPV lower head. The temperature and stress fields of the vessel wall, and the plasticity and creep deformation of the lower head are calculated. Combining the plasticity and creep rupture criteria, the failure was analyzed. Results show that the deformation of the structure will be greatly increased when creep is considered. During the IVR strategy under severe accident, the main failure mode of the RPV lower head is creep failure instead of plastic failure. Further analysis implies that the internal pressure has a significant influence on the creep deformation and failure time. This paper provides the method for the creep and failure analysis of the RPV lower head under severe accident. (authors)

[en] This paper studied the recent progress research on the reactor pressure vessel (RPV) under severe accident at home and abroad, including the research results of the creep failure of reactor pressure vessel, which provided a great number of test results, creep constitutive model, creep damage theory, failure criterion, finite element method and some related high temperature material parameters. The summarized results can be particularly useful for conducting the analysis and experimental research on creep failure of RPV under severe accident. (authors)

[en] The in vessel retention (IVR) strategy is an important mitigation measure for serious accidents in nuclear power plants. In this paper, the finite element method is used to study the mechanical behavior of the reactor pressure vessel (RPV) lower head under the action of the melt during the IVR strategy. According to the thermal and mechanical loads distribution characteristics that the core melt transferred to the pressure vessel wall, the temperature field and stress field distribution of the lower head were obtained by calculation, and the effects of thermal expansion, internal pressure and other loads on the mechanical response of the structure were investigated. The comparison between the elastic and elasto-plastic behavior of materials was conducted as well. Results show that the stress and deformation caused by thermal expansion are much larger than those of the vessel dead weight, molten pool pressure and cooling water pressure;when the internal pressure is greater than 1 MPa, it has a significant effect on the mechanical response of the structure;under the action of the core melt, the structure will produce certain plastic strain that cannot be ignored, and the elasto-plastic analysis method is more reasonable. (authors)

[en] Producing through-wall crack specimen of nuclear pipe is necessary for the LBB research, this paper carried out numerical simulation on the crack growth by finite element method and test verification through four point bending test, including the design of main components of four point bending facility and crack growth tests. The numerical calculation and test results are compared, which shows that the loading cycles of pipe through-wall crack growth are in accordance with the numerical results. Numerical method can simulate the process of making through-wall crack specimen precisely, which is a useful reference for the crack producing of LBB pipe. (authors)

[en] With the increase of coolant outlet temperature in advanced nuclear reactors, it is necessary to study the elevated temperature property of the loop piping material. In this paper, the elevated temperature mechanical property of three common materials used in advanced reactor, i.e., type 304 stainless steel, type 316 stainless steel and modified 9Cr-1Mo steel (grade 91, T91, P91 steel), have been researched based on standards and literatures. The chemical composition, tensile and creep property of the three kinds of materials at elevated temperature has been investigated. Results show that the tensile strength of grade 91 steel is basically higher than that of type 304 and 316 stainless steel at elevated temperature, while its creep strength is lower than type 316 stainless steel, but higher than type 304 stainless steel. The research results can provide some reference for the choice of loop piping material for generation IV reactors (e.g., supercritical water cooled reactor, sodium-cooled fast reactor and lead-cooled fast reactor) and other advanced reactors (e.g., traveling wave reactor). (authors)

[en] The heat treatment of SA508-3 steel was carried out by means of quenching and manual aging. The microstructure and mechanical properties of the prepared SA508-3 steel were analyzed and tested. The results show that the main composition of the matrix of SA508-3 steel is bainite, and the main composition of the second phase is alloy cementite containing Fe, Mn and C. The second phase is distributed at both the matrix and grain boundaries. The second phase can prevent the dislocation from moving by fixing the dislocation, and the second phase at the grain boundary can strengthen the matrix by hindering the grain boundary movement. The stress-strain curves of SA508-3 steel under different loading rates show that when the strain rate is greater than 0.5 m s${}^{-<!--\; ???\; -->1}$, the fracture mode of the steel is brittle fracture, and when the strain rate is less than 0.5 m s${}^{-<!--\; ???\; -->1}$, the fracture mode of the steel is ductile-brittle bonding fracture. The second phase of the crack first diffuses to the grain boundary, reducing the strength of the grain boundary. When the loading rate is high, the second phase at the grain boundary cannot diffuse in time, and the material undergoes transgranular fracture and intergranular fracture.