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The fatigue tests of Chinese RPV steel (A508-3) were carried out under the simulated primary coolant environment of AP1000 which is a kind of 3rd generation nuclear power plant. The effects of temperature, strain rate and dissolved oxygen (DO) on fatigue life were analyzed. During the fatigue test, the material exhibited a cyclic hardening followed by cyclic softening behavior, with no obvious cyclic stability region. In logarithmic coordinates, the S–N curve showed a bilinear relationship. As the strain amplitude increased from 0.2 to 0.6%, the fatigue life approximately decreased from 105 to 3 × 102. The environmental correction factor (Fen), which is defined as the ratio of life in air at room temperature to that in the primary coolant environment, decreases as the temperature decreases from 321 to 150 °C. The Fen also decreases as the DO decreases from 0.6 to 0.03 ppm while it increases as the strain rate decreases from 1 to 0.0004%/s. Compared with the NUREG6909 model, the presented model is more applicable for fatigue life prediction of Chinese A508-3 steel. [ABSTRACT FROM AUTHOR]

• SiC crystal was used as an irradiation temperature monitor in a thermal reactor. • The irradiation temperature is overestimated by an electrical measurement of SiC. • Annealing-induced elimination of C i accounts for the electrical changes in SiC. • Hardness of SiC increases with neutron irradiation and then decreases by annealing. Chemical vapor deposited polycrystalline SiC was irradiated with neutrons in a thermal reactor at nominal 550 ℃ to 2 × 1019 n/cm2 (E > 0.1 MeV). Isochronal annealing of the irradiated SiC was performed to determine the sample irradiation temperature, which is based on the annealing temperature dependence of the sample electrical resistivity. The estimated irradiation temperature of 615 ℃ is significantly higher than the target temperature recorded by thermocouple. The Raman spectroscopy results indicate that the C-C bonds formed in the irradiated SiC are preferentially annealed out than Si-Si bonds. The behavior is attributed to the fast migration and elimination of interstitial C during annealing, which accounts for the annealing-induced electrical resistivity changes in SiC. The hardness of SiC characterized by nanoindentation is found to increase with neutron irradiation and decrease after post-irradiation annealing, which is associated with the production and recovery of radiation defects in SiC. [ABSTRACT FROM AUTHOR]

LT21 a type of aluminum alloy used for the irradiation channel of the first heavy water research reactor (HWRR) in China. Studying the mechanical property of irradiated LT21 aluminum under actual service conditions is essential for evaluating its application property. In this paper, tensile specimens of irradiated LT21 were manufactured from the decommissioned irradiation channel of an HWRR; then, tensile tests were carried out, and then the fracture surfaces were observed. The effect of neutron irradiation on tensile behavior and the failure mechanism was analyzed by comparing the result of irradiated and unirradiated LT21 specimens. The results show that, with the thermal neutron flux increasing to 2.38 × 1022 n/cm2, the YS gradually increased from the initial 158 MPa to 251 MPa, the UTS increased from 262 MPa to 321 MPa, and the elongation decreased from 28.8% to about 14.3%; the brittle fracture of the LT21 specimen appeared after irradiation, and the proportion of brittle fracture increased as the neutron fluence increased; the nanophase structures, with a size of less than 50 nm, were precipitated in the LT21 aluminum alloy after neutron irradiation. Transmutation Si is presumed to be the main cause of the radiation effect mechanism of LT21. [ABSTRACT FROM AUTHOR]

Reactor pressure vessel (RPV) is the critical un-changeable component of the reactor during its service lifetime, which prevents the radioactive leak of the nuclear power plant core. The irradiation test (about 10 × 10 19 cm − 2 , E > 1 MeV) in research reactor of the pressure vessel material was carried out, and the charpy impact test has been carried out before and after the neutron irradiation. Analysis of the impact energy and the fracture morphology has been done to estimate the embrittlement due to neutron irradiation. It shows that the main effects of neutron irradiation to fracture are the crack initiation and stable expansion process. And there also are a small amount of intergranular fracture in the unstable crack expansion after neutron irradiated which aware us pay more attention to the increasing intergranular fracture behavior of higher neutron fluence level after 60a nuclear power plant operation. [ABSTRACT FROM AUTHOR]

• The K IC and T 0 of the RPV material were calculated using impact specimens. • K IC was obtained form impact specimen's fracture, F-t curve, size, shape factor. • T 0 was obtained by the multi-temperature formula applicable to the impact test. • K IC , T 0 are consistent with results obtained by CTt according to the ASTM standard. Accurately realizing the use of Charpy impact test to measure fracture toughness and determine reference temperature of low alloyed steel with ferritic microstructure has important practical engineering significance for the structural integrity assessment of reactor pressure vessels. In this paper, a methodology for measuring fracture toughness and determining reference temperature for RPV steels by Charpy impact test was proposed. The method referred to the ASTM E1820 and E1921 standard. By analyzing the physical meaning of the Charpy impact specimen's fracture, load-time curve and size parameters, the fracture toughness of the material was directly calculated according to the linear elastic fracture mechanics. The reference temperature was calculated using the multi-temperature procedure formula. The reference temperature and fracture toughness were corrected according to two factors, such as the difference in fracture toughness caused by the difference in the loading rate of the stress intensity factor and the specimen shape and size between Charpy impact test and compact tension test. And the validity of the fracture toughness and reference temperature was evaluated. The corrected two are basically consistent with the data obtained by the compact tension test in accordance with ASTM E1820 and E1921 standards. This method may be used to solve the problem that the nuclear power plant needs to use the impact test data to evaluate the fracture toughness and reference temperature of the reactor pressure vessel in the structural integrity assessment of the reactor pressure vessel operation and life extension process. [ABSTRACT FROM AUTHOR]