Results 1 - 10 of 17483
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[en] Negative creep using precision dilatation experiments is investigated on a broad variety of single, dual and multiphase nickel-based superalloys. Pure nickel and dilute binary nickel-based alloys show no signs of negative creep. However, with higher contents of Cr and Al and in highly alloyed ternary and multicomponent nickel-based alloys negative creep is observed. Short range ordering of Cr and/ or Al are identified to cause negative creep at 550 °C. Carbon additions leading to retarded carbide precipitation or transformation can enhance negative creep.
[en] According to the microstructural evolution during longterm thermal exposure at 1100 °C, the creep rupture life of Ni-based single crystal superalloys at 980 °C/270 MPa was evaluated. The microstructure was characterized by means of scanning electron microscopy, X-ray diffraction and related image processing methods. The size of γ’ precipitates and the precipitation amount of topologically close-packed increased with the increase in thermal exposure time, and coarsening of the γ’ precipitates led to the simultaneous increase of the matrix channel width. The relationship between the creep rupture life and the lattice misfit of γ/γ’, the coarsening of γ’ precipitate and the precipitation of TCP phase are systematically discussed. In addition, according to the correlation between γ’ phase evolution and creep characteristics during thermal exposure, a physical model is established to predict the remaining creep life.
[en] Present work is in continuation of the earlier reported work on high temperature tensile and creep-stress rupture properties of Calandria material SS 304L that are required for severe accident analysis. The report covers the observations from fractography and metallography studies of crept specimens. Activation energy of creep of SS 304L has been evaluated and creep behavior has been studied in conjunction with available deformation mechanism maps from the literature. The results would be helpful in understanding and qualitative assessment of structural degradation of Calandria in case of a beyond design basis severe accident in PHWRs. (author)
[en] Herein the tensile creep behavior and microstructure evolution of an as-cast Mg-9.82Gd-0.38Zr alloy under different stresses at 250 °C are investigated. The results of creep test show that the creep strain and steady creep rate increase with the creep stress at 250 °C. The fitted stress exponent n value (4.4) and transmission electron microscopy analysis suggest that the steady creep is dominated by the dislocation climb and glide mechanisms together. At the primary stage, due to the segregation of the Gd content, a fine-strengthening β′ phase precipitates near the α-Mg grain boundary. Then the transformation of β′ precipitates to a needle-like β phase, coarsening, and growth of the β′ phase toward the intragranular direction gradually deteriorate the creep resistance of the experimental alloy. Moreover, two kinds of precipitate-free zones (PFZs) appear at the end of the secondary creep stage and the PFZs nearly perpendicular to the applied stress direction widen gradually because of the directional diffusion. The nucleation and propagation of creep crack along the cracked eutectic phase and PFZs lead to the greatly rising creep rate at the end of the tertiary creep stage, which eventually results in the intergranular creep fracture of the experimental alloy. (© 2021 Wiley-VCH GmbH)
[en] This paper describes benchmark analysis of independently programmed structural reliability evaluation codes, REAL-P and GENPEP. An upper core structure of a prototype fast breeder reactor in Japan, MONJU, was chosen, and crack initiation time and crack propagation due to fatigue-creep interaction damage was evaluated in deterministic and probabilistic manners. Evaluation procedures follow the new guidelines on reliability evaluation of fast reactor components issued by JSME. The results obtained by two codes were compared, and the effects of differences in treatments of which details are not prescribed in the guidelines on results were discussed. As result, although slight difference was recognized in crack initiation evaluation especially due to difference in fairing treatment of fatigue life curves, the results estimated by two codes generally agreed very well for both deterministic and probabilistic evaluations. It was shown that the effects of differences in treatments of which details are not prescribed in the guidelines on results are small for structural reliability evaluation of fatigue-creep interaction damage, which is one of typical degradation mechanisms for fast reactor passive components. (author)
[en] Accident Tolerant Fuels have been widely studied since the Fukushima-Daiichi accident in 2011 as one of the options on how to further enhance the safety of nuclear power plants. Deposition of protective coatings on nuclear fuel claddings has been considered as a near-term concept that will reduce the high-temperature oxidation rate and enhance accidental tolerance of the cladding while providing additional benefits during normal operation and transients. This study focuses on experimental testing of Zr-based alloys coated with Cr-based coatings using Physical Vapour Deposition. The results of long-term corrosion tests, as well as tests simulating postulated accidents, are presented. Zr–1%Nb alloy used as nuclear fuel cladding serves as a substrate and Cr, CrN, CrxNy layers are deposited by unbalanced magnetron sputtering and reactive magnetron sputtering. The deposition procedures are optimized in order to improve coating properties. Coated as well as reference uncoated samples were experimentally tested. The presented results include standard long-term corrosion tests at 360∘C in WWER water chemistry, burst (creep) tests and mainly single and double-sided high-temperature steam oxidation tests between 1000 and 1400∘C related to postulated Loss-of-coolant accident and Design extension conditions. Coated and reference samples were characterized pre- and post-testing using mechanical testing (microhardness, ring compression test), Thermal Evolved Gas Analysis analysis (hydrogen, oxygen concentration), optical microscopy, scanning electron microscopy (EDS, WDS, EBSD) and X-ray diffraction.
[en] As a candidate of the accident tolerant fuel (ATF) in the KAERI, the microcell UO2 pellet and the surface modified cladding with coating are being developed as the near term technology. The microcell UO2 pellet is to enhance the fission product retention and to increase thermal conductivity. The surface modified cladding is based on the coating technology on the conventional Zirconium alloy cladding. It is to enhance the oxidation and deformation resistance of the fuel cladding. In the IAEA-CRP ACTOF, the research objectives and anticipated outcomes of KAERI were the development and implementation in the FRAPCON/FRAPTRAN code of ATF models for the coated cladding and metallic microcell UO2 pellet of KAERI.In order to evaluate the mechanical behaviours of the ATF cladding under the normal operation conditions, a new analytical module, FRACAS-CT, was developed based on the thick-wall theory to consider the multi-layered structure of the coated cladding. The FRACAS-CT model was verified by comparison with an equivalent finite element model. And the module was implemented into FRAPCON code with consideration of creep and stress relaxation behaviours of the multi-layered cladding. The implemented FRACAS-CT can simulate the mechanical response and fuel performance of the multi-layered ATF cladding. The preliminary analysis of the fuel performance for the KAERI’s ATF concept was summarized. The major material properties of ATF such as the thermal conductivity and thermal expansion of the pellet and the corrosion behaviour of the CrAl-coated cladding were modified based on out-of-pile test results. The differences compared to a conventional UO2-Zircaloy fuel was assessed. From the FRAPCON results under the normal operation condition, ATF shows a significant advantage in the reduction of the fuel centreline temperature, cladding oxidation thickness, fission gas release, and so on, because of the increased thermal conductivity of the metallic microcell pellet and the oxidation resistance of the CrAl-coated cladding. (author)
[en] The creep behavior of Ni–Cr–W alloy at 950 °C has been investigated by a novel creep testing system which is capable of in-situ measurement of strain. Tubular specimens were pressurized with argon gas for effective stresses up to 32 MPa. Experimental results show that the thermal fatigue reduces the creep life of the tubular specimens and with the introduction of thermal cycling fatigue the primary stage disappears and the creep rate higher than the pure thermal creep (without thermal fatigue). Also the creep behavior of Ni–Cr–W alloy doesn't consist in the secondary stage. A new creep equation has been derived and implemented into finite element method. The results from the finite element analyses are in good agreement with the creep experiment
[en] Several models were integrated to the DIONISIO code within the framework of the IAEA Research Project “Fuel Modeling in Accident Conditions (FUMAC)”, to take account of accidental conditions, in particular the loss of coolant accidents (LOCA). A specially designed thermal-hydraulic subroutine provides a simplified description of the rod environment in normal or accidental conditions. The heat transfer coefficients that account for the different coolant regimes, in single or double phases, are activated as the corresponding conditions occur. The simulation of a considerable number of experiments has shown that, despite its simplicity this subroutine gives adequate predictions of the conditions in a vertical cooling channel, quite similar to those given by the thermal-hydraulic codes. The description of the fuel rod atmosphere is improved with the incorporation of this subroutine since it provides fairly realistic boundary conditions for the simulation of the fuel rod behavior, without requiring the intervention of external specific codes. Models of high temperature oxide growth (ZrO2) and hydrogen capture and release by the cladding in steam were also included. Moreover, the model of cladding creep predicts the conditions for ballooning and eventually, those for catastrophic failure (burst) and its localization. The calculation scheme makes a partition of the rod length into a number of segments defined by the user. In each segment the local conditions are considered to calculate, with the synchronous work of all the subroutines, the physical and chemical parameters in one representative pellet. Then, a description of the whole rod is obtained by coupling all the segments. This strategy has yielded accurate simulations of a wide variety of cases, either in normal or LOCA type conditions. Exhaustive comparisons were carried out with several thermal-hydraulic codes (COBRA-IV, RELAP5-Mod3.1, SOCRAT, ATHLET-Mod 1.1) and with a number of experiments like those of the IFA–650 series (-1,-2,-9,-10,-11), PUZRY, QUENCH-L0/L1 (for which a new working scheme was specially developed in DIONISIO), CORA-15, IAEA-SPE-4, among others. (author)
[en] The paper illustrates the results of the computer assessment of the form alteration in WWER-1000 core baffle obtained via the solution to the coupled thermoelastoplastic task considering the strains of irradiation growth and creep. In the modeling of the contact conditions, the temperature redistribution is considered due to the incompliance of the cool-ant flow in the contact zone between the core baffle and in-vessel core barrel with the design conditions. The modern approaches to the modeling of strains of the irradiation growth and irradiation creep in austenite steels are used in the space-limited environment under neutron exposure and elevated temperature. The finite element analysis involves the mixed scheme of the finite element method, which allows determination of the stress-strain state with high accuracy. The calculations are performed in the two-dimensional statement for the cross-section of the core baffle with the maximum damaging dose and irradiation temperature under the condition of the generalized plane strain. The results of the calculations are presented for full-scale reactor operation and scheduled shutdown to recharge the fuel cluster at the end of core life. The data on the distribution and value of the gap between the core baffle and barrel, as well as the spacer grids of the edge fuel assemblies and reactor core baffle edges, have been obtained from the median values of the dose dependence on swelling at different temperatures in Kh18N10T austenite steel. (author)