Results of irradiated cladding tests and clad plate experiments (original) (raw)
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Crack Growth Rate and Fracture Toughness Tests on Irradiated Cast Stainless Steels
2013
Cast austenitic stainless steels are used in the cooling system of light water reactors for components with complex shapes, such as pump casings, valve bodies, and coolant piping. In the present study, crack growth rate and fracture toughness JR curve tests were performed on irradiated cast stainless steels in low-corrosion-potential environments (low-dissolved-oxygen high-purity water or simulated pressurized water reactor environment) at 320°C. Both asreceived and thermally aged materials were included to investigate the combined effect of thermal aging and irradiation embrittlement on the fracture behavior of cast stainless steels. The samples were irradiated to approximately 0.08 dpa at the Halden reactor. Good resistance to corrosion fatigue and stress corrosion cracking was observed for all samples. Thermal aging had little effect on the crack growth behavior at this dose level. Cleavage-like fracture was the dominant cracking morphology during the crack growth rate tests, and the ferrite phase was deformed to a lesser extent compared with the surrounding austenite phase. The fracture toughness results showed a dominant effect of neutron irradiation, and the fracture resistances were decreased considerably for all cast specimens regardless of their thermal aging conditions. The reduction in fracture toughness was more significant in the unaged than thermally aged materials. Nonetheless, the fracture toughness values of thermally aged specimens were about 30% lower than their unaged counterparts, suggesting a combined effect of thermal aging and neutron irradiation in cast stainless steel.
2015
Cast austenitic stainless steels are used for components with complex shape in the cooling system of light water reactors. Due to both thermal aging and irradiation embrittlement, the long-term performance of CASS materials is of concern. To assess the extent of embrittlement, crack growth rate and fracture toughness J-R curve tests were performed in this study on thermally aged and unaged CF-3 at ~320°C in simulated LWR coolant environments. The samples were miniature compact tension specimens irradiated to a fast neutron fluence of 5x10 n/cm (E>1 MeV) or 0.08 dpa. While environmentally enhanced cracking was observed under cyclic loading, SCC crack growth rates were low for these CF-3 specimens. Following the crack growth rate tests, fracture toughness J-R curve tests were carried out in the same test environments. The irradiation-induced reduction in fracture toughness was more evident in the unaged than aged specimens. At 0.08 dpa, the fracture toughness values of unaged speci...
Journal of Nuclear Materials, 2007
Irradiation embrittlement studies rely very often on Charpy impact data, in particular the ductile-to-brittle transition temperature (DBTT). However, while the DBTT-shift is equivalent to the increase of the fracture toughness transition temperature of ferritic steels, it is not the case for ferritic/martensitic steels. The aim of this study is to critically assess experimental data obtained on a 9%Cr-ferritic/martensitic steel, Eurofer-97, to better understand the underlying mechanisms involved during the fracture process. More specifically, a dedicated analysis using the load diagram approach allows to unambiguously reveal the actual effects of irradiation on physically rather than empirically based parameters. A comparison is made between a ferritic and ferritic/martensitic steel to better identify the possible similarities and differences. Tensile, Charpy impact and fracture toughness tests data are examined in a global approach to assess the actual rather than apparent irradiation effects. The adequacy or inadequacy of the Charpy impact test to monitor irradiation effects is extensively discussed.
Charpy impact properties of martensitic 10.6% Cr steel (MANET-I) before and after neutron exposure
Fusion Engineering and Design, 1995
An extensive irradiation programme (FRUST/SIENA) was elaborated to study the influence of radiation upon the Charpy impact characteristics of the MANET-I martensitic 10.6% Cr steel. In addition to unirradiated reference specimens, 87 irradiated subsize Charpy specimens (3 x 4 x 27 mm') were examined under eight different heat treatments at irradiation temperatures between 290 and 475 "C and exposure doses of 5-15 dpa. A newly developed instrumented Charpy impact bending test and evaluation system was used in the tests. The impact properties characterized by the upper shelf energy (USE) and the transition temperature from brittle to ductile material behaviour (DBTT) were determined before and after irradiation. Clear empirical correlations could be established according to which irradiation at lower temperatures provokes a shift in DBTT and USE independent of the initial material state with dose saturation at lo-15 dpa. An elevated irradiation temperature alleviates the irradiationinduced material deterioration.
2013
Slow-strain-rate tensile (SSRT) tests have been performed on irradiated specimens in a simulated pressurized water reactor (PWR) environment. The samples are miniature tensile specimens of various austenitic stainless steels (SSs) with different thermal-mechanical treatments commonly used for reactor core internal components. Neutron irradiation was carried out in the BOR-60 reactor, a sodium-cooled, fast breeder reactor in Russia, at ~320°C. The damage doses of the specimens are 5, 10, and 48 dpa (displacements per atom). All irradiated materials show significant irradiation hardening and loss of ductility in the SSRT tests. The yield strengths of cold-worked are higher than that of solution-annealed samples at all doses up to 48 dpa. While the irradiation hardening seems to saturate between ~5 and ~10 dpa, the loss of ductility continues to increase above 10 dpa. Strain softening is also observed for all irradiated materials above 5 dpa. Fractographic examinations show that ductile dimple fracture is the dominant morphology for all SSRT tests in the PWR environment. Small areas of transgranular, mixedmode and cleavage fractures are seen on most fracture surfaces in PWR water tests. Intergranular cracking is also observed in 48-dpa Types 316 and 347 SSs. Cracking susceptibility of the tested materials was evaluated with fracture morphology and time to failure. In general, high-doses cold-worked SSs are more susceptible to transgranular cleavage cracking in the PWR environment. Solution-annealed Type 347 SS is susceptible to intergranular cracking at 45 dpa in the PWR environment.
Engineering Failure Analysis, 2019
The mechanical performance of reactor pressure vessel (RPV) materials is an important factor in determining the safety and economics of the operation of a nuclear power plant. The ductile-tobrittle transition temperature (DBTT) tested by Charpy impact test is an important parameter for evaluating the RPV embrittlement. In this paper, the Charpy impact test of RPV steel prior and after irradiation was carried out at different temperatures. The fracture morphology was observed by scanning electron microscopy. The irradiation effect on the impact fracture behavior was analyzed in the DBTT temperature zone. Results show that in the DBTT range, the crack initiation and propagation energy are obviously reduced for the irradiated material. After irradiation, the dimple area on the fracture surface is significantly reduced, and the distance from the crack initiation source to the gap is decreased. For the RPV steel with low Cu steel content, matrix damage becomes the main factor on material irradiation embrittlement.
Busby/Environmental Degradation, 2012
Irradiation-assisted stress corrosion cracking is of concern for the safe and economic operation of light water reactors. In this study, cracking susceptibility of austenitic stainless steels was investigated by using slow strain rate tensile (SSRT) tests in a simulated pressurized water reactor (PWR) environment. The specimens were irradiated to 5, 10, and 48 dpa in the BOR60 reactor at 320°C. The SSRT results showed that yield strength was increased significantly in irradiated specimens while ductility and strain hardening capability were decreased. Irradiation hardening was found to be saturated below 10 dpa. The irradiated yield strength of cold-worked specimens was higher than that of solution-annealed specimens. Fractographic examinations were also performed on the tested specimens, and the dominant fracture morphology was ductile dimples. Intergranular cracking was rarely seen on the fracture surface. Transgranular cleavage cracking, however, was found more frequently on the specimen tested in simulated PWR environment. * SA = solution annealed, CW = cold worked, HP = high purity. 2.2 Irradiations All specimens were irradiated in BOR-60, a sodium-cooled fast breeder reactor located in the Research Institute of Atomic Reactors (RIAR), Dimitrovgrad, Russia. Irradiations were performed in two experiments, Boris-6 and-7, and at three displacement dose levels (5, 10, and 48 dpa) [14]. Neutron dosimeters were loaded in the irradiation rig along with the specimens, and the final dosimetry was carried out by RIAR after irradiation. During the irradiation experiments, the tensile specimens were separated in bundles (four specimens in each bundle) and were in contact with sodium coolant. The irradiation temperature was controlled by
Tensile and Charpy Impact Properties of Irradiated Reduced Activation Ferritic Steels
Effects of Radiation on Materials: 18th International Symposium, 1999
Tensile tests were conducted on eight reduced-activation Cr-W steels after irradiation to 15-17 and 26-29 dpa, and Charpy impact tests were conducted on the steels irradiated to 26-29 dpa. Irradiation was in the Fast Flux Test Facility at 365°C on steels containing 2.25-12% Cr, varying amounts of W, V, and Ta, and O.l%C. Previously, tensile specimens were irradiated to 6-8 dpa and Charpy specimens to 6-8, 15-17, and 20-24 dpa. Tensile and Charpy specimens were also thermally aged to 20000 h at 365°C. Thermal aging had little effect on the tensile behavior or the ductile-brittle transition temperature (DBTT), but several steels showed a slight increase in the upper-shelf energy (USE). After =7 dpa, the strength of the steels increased (hardened) and then remained relatively unchanged through 26-29 dpa (Le., the strength saturated with fluence). Postirradiation Charpy impact tests after 26-29 dpa showed that the loss of impact toughness, as measured by an increase in DBTT and a decrease in the USE, remained relatively unchanged from the values after 20-24 dpa, which had been relatively unchanged from the earlier irradiations. As before, the two 9Cr steels had the most irradiation resistance.
Tensile properties of 9Cr-1MoVNb and 12Cr-1MoVW steels irradiated to 23 dpa at 390 to 550 ° C
Journal of Nuclear Materials, 1991
Normalized-and-tempered 9Cr-1MoVNb and 12Cr-1MoVW steels were irradiated in the Experimental Breeder Reactor II (EBR-II) at 390,450,500, and 550°C to displacement damage levels of up to 23 to 25 dpa. Tensile tests were conducted at the irradiation temperatures on three types of specimens: irradiated specimens, normalized-and-tempered specimens, and specimens thermally aged 5000 h at the irradiation temperatures. Observations from these tests were compared with results on these same materials irradiated in EBR-II at the same temperatures to 9 to 13 dpa.
Damage assessment in structural metallic materials for advanced nuclear plants
Journal of Materials Science, 2010
Future advanced nuclear plants are considered to operate as cogeneration plants for electricity and heat. Metals and alloys will be the main portion of structural materials employed (including fuel claddings). Due to the operating conditions these materials are exposed to damaging conditions like creep, fatigue, irradiation and its combinations. The paper uses the most important alloys: ferritic-martensitic steels, superalloys, oxide dispersion strengthened steels and to some extent titanium aluminides to discuss its responses to these exposure conditions. Extrapolation of stress rupture data, creep strain, swelling, irradiation creep and creep-fatigue interactions are considered. Although the stress rupture-and the creep behavior seem to meet expectations, the long design lives of 60 years are really challenging for extrapolations and particularly questions like negligible creep or occurrence of diffusion creep need special attention. Ferritic matrices (including oxide dispersion strengthened (ODS), steels) have better irradiation swelling behavior than austenites. Presence and size of dispersoids having a strong influence on high-temperature strength bring only insignificant improvements in irradiation creep. A strain-range-separation based approach for creep-fatigue interactions is presented which allows a real prediction of creep-fatigue lives. An assessment of capabilities and limitations of advanced materials modeling tools with respect to damage development is given.