Assessment of creep in reactor-irradiated CuCrZr alloy intended for the ITER first wall panels (original) (raw)
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Journal of Nuclear Materials, 2011
Precipitation hardened CuCrZr alloy is a promising heat sink and functional material for various applications in ITER, for example the first wall, blanket electrical attachment, divertor, and heating systems. Three types of thermo-mechanical treatment were identified as most promising for the various applications in ITER: solution annealing, cold working and ageing; solution annealing and ageing; solution annealing and ageing at non-optimal condition due to specific manufacturing processes for engineering-scale components. The available data for these three types of treatments were assessed and minimum tensile properties were determined based on recommendation of Structural Design Criteria for the ITER In-vessel Components. The available data for these heat treatments were analyzed for assessment of neutron irradiation effect. Using the definitions of the ITER Structural Design Criteria the design allowable stress intensity values are proposed for CuCrZr alloy after various heat treatments.
Title: Final Report on In-reactor Creep-fatigue Deformation Behaviour of a CuCrZr Alloy: COFAT 1
2000
At present, practically nothing is known about the deformation behaviour of materials subjected simultaneously to external cyclic force and neutron irradiation. The main objective of the present work is to determine experimentally the mechanical response and resulting microstructural changes in CuCrZr(HT1) alloy exposed concurrently to flux of neutrons and creep-fatigue cyclic loading directly in a fission reactor. Special experimental facilities
Effect of irradiation dose on mechanical properties and fracture character of Cu//SS joints for ITER
Journal of Nuclear Materials, 2007
Cu//SS type joints are essential for the heat-sink systems of the ITER high-heat flux components. A number of technologies have been proposed for the production of such bimetallic structures, including brazing, friction welding, HIP and cast-copper-to-steel. In this paper, the authors present the results of investigations into the irradiation resistance of Glid-CopAl25//316L(N) and CuCrZr//316L(N)-type joints produced by the HIP and cast-copper-to-steel (CC) technologies. Specimens of the joints were irradiated in the RBT-6 reactor in the dose range of 10 À3-10 À1 dpa at T irr = 150°C. Irradiation causes strengthening of the joint specimens and the uniform elongation drops from 7% in the initial state to 1-2%. However, the total elongation remains at the relatively high level of $10%. The investigations performed make it possible to recommend joints of CuCrZr//316L(N) (CC) and CuCrZr//316L(N) (HIP) type produced by the cast-copper-to-steel and HIP technologies, respectively, for ITER applications.
Irradiation Testing of Structural Materials in Fast Breeder Test Reactor
Fast Breeder Test Reactor (FBTR) at Kalpakkam, India is a sodium cooled fast reactor with neutron flux level of the order of 10 15 n/cm 2 /s and temperature of coolant in the range of 600-720K (330-450°C), which is being used for the development of fuel and structural materials required for Indian Fast Reactor Programme. Irradiation performance testing on structural materials is being carried out by subjecting prefabricated specimens to desired experimental conditions as part of planned irradiation experiments, and by testing of material samples sourced from actual fuel clad tubes / fuel assembly wrapper tubes irradiated to various fuel burn up levels in FBTR. Pressurised capsules of Zirconium alloys and D9 alloy (modified stainless steel type 316 with controlled additions of titanium and silicon) have been developed to determine the in-reactor creep performance of indigenously developed zirconium alloys and D9 alloy. Pressurised capsules made of zirconium alloys were subjected to fluence levels up to 1.1 x 10 21 n/cm 2 (E> 1 MeV) in FBTR at temperatures of 579 to 592K and diameter measurements were carried out in the hot cell facility to determine the irradiation creep rate. Pressurised capsules of D9 alloy are currently undergoing irradiation at a temperature of 623K in FBTR along with small size tensile test specimens and shear punch test specimens of D9. Non-instrumented gas-gap type irradiation capsule has been developed to achieve higher irradiation temperatures (673 to 873K) of structural material specimens. The irradiation induced mechanical property changes in cold worked AISI Type SS316 fuel cladding of FBTR have been determined from tensile testing of portions of irradiated fuel clad tubes in the hot cells. Tests were carried out on clad tubes with dpa ranging from 13 to 83 at various test temperatures from ambient (300K) to irradiation temperature (790K). Shear punch tests have been used for characterizing the tensile property changes in cold worked AISI Type SS 316 wrapper material of FBTR fuel assemblies. From the results of shear punch tests on irradiated specimens, using correlation equations, the tensile properties of the wrapper material irradiated to various dpa ranging from 30 to 83 have been estimated. A considerable increase in the strength and decrease in the ductility of the wrapper material with increasing dpa was observed from the results. This paper discusses the salient features of irradiation facilities available at FBTR, irradiation experiments carried out on structural materials, and some of the important results obtained from tests on irradiated structural materials.
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
Mechanical Properties of Advanced Gas-Cooled Reactor Stainless Steel Cladding After Irradiation
Journal of Materials Engineering and Performance
The production of helium bubbles in advanced gas-cooled reactor (AGR) cladding could represent a significant hazard for both the mechanical stability and long-term storage of such materials. However, the high radioactivity of AGR cladding after operation presents a significant barrier to the scientific study of the mechanical properties of helium incorporation, said cladding typically being analyzed in industrial hot cells. An alternative non-active approach is to implant He 2+ into unused AGR cladding material via an accelerator. Here, a feasibility study of such a process, using sequential implantations of helium in AGR cladding steel with decreasing energy is carried out to mimic the buildup of He (e.g., 50 appm) that would occur for in-reactor AGR clad in layers of the order of 10 lm in depth, is described. The implanted sample is subsequently analyzed by scanning electron microscopy, nanoindentation, atomic force and ultrasonic force microscopies. As expected, the irradiated zones were affected by implantation damage (< 1 dpa). Nonetheless, such zones undergo only nanoscopic swelling and a small hardness increase ($ 10%), with no appreciable decrease in fracture strength. Thus, for this fluence and applied conditions, the integrity of the steel cladding is retained despite He 2+ implantation.
Journal of Nuclear Materials, 2018
Post-irradiation examination (PIE) in the form of transmission electron microscopy (TEM) was used to characterize the microstructure of several specimens of Zircaloy-2 and Zircaloy-2 plus 1% Nb that had previously underwent in-pile creep testing in the HALDEN reactor. The purpose of the examination was to explore a microstructural basis for an apparent substantially increased rate of hardening over that observed with similar materials in the higher fast-flux environments of the ATR or HFIR, and to investigate the increased creep strength exhibited by the Nb-containing alloy relative to pure Zircaloy-2. The analysis of irradiation-induced defects indicated a higher than expected density, which was consistent with the observed high hardening rate. Modeling based on mean field rate theory suggests the lower neutron flux in the HALDEN reactor results in a higher fraction of irradiation-induced defects being available for sink (loop) nucleation and growth.
Nuclear Engineering and Design, 2014
Irradiation stress corrosion cracking (IASCC) is a critical degradation mechanism for reactor internal components that contributes to the safety and economic operation of reactors. As nuclear power plants age and irradiation dose increases, IASCC becomes an increasingly important issue because of its potential impact on the integrity of reactor internal components. In this study, slow strain rate tensile tests were conducted on irradiated tensile specimens at strain rates between 3 to 7x10-7 s-1 to evaluate cracking susceptibility of austenitic stainless steels in simulated light water reactor (LWR) environments. Significant increases in yield strength were observed for all irradiated specimens and a dose dependence of irradiation hardening was obtained at temperatures relevant to LWRs. Ductility and strain hardening capability were also found decrease rapidly with the increase of dose. After the tests, the specimens were examined using a scanning electron microscopy to characterize fracture morphology. The area fractions of non-ductility fracture were used to evaluate the IASCC susceptibility along with the tensile properties. IASCC susceptibility was also compared for several stainless steels irradiated in the Halden and BOR-60 reactors. A possible neutron spectrum effect was discussed.