Irradiation Testing of Structural Materials in Fast Breeder Test Reactor (original) (raw)
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Performance Assessment of Fuel and Core Structural Materials Irradiated in FBTR
Energy Procedia, 2011
Post-irradiation examination (PIE) is a vital link in the nuclear fuel cycle for providing valuable feedback on the performance and residual life of the fuel and structural materials to designers, fabricators, and reactor operating personnel. The challenging task of setting up of α,β,γ inert atmosphere hot cell facility for PIE of Fast Breeder Test Reactor (FBTR) was accomplished successfully and irradiation performance of the FBTR mixed carbide fuel was assessed stage wise at various burnups starting from 25 GWd/t upto 155 GWd/t. With FBTR being used as a test bed for irradiation experiments on various FBR fuels and structural materials, PIE of various materials subjected to experimental irradiation like the PFBR MOX fuel, FBTR grid plate material have also been carried out to provide valuable feedback to the designers. This paper highlights the (i) results of comprehensive PIE carried on mixed carbide fuel & structural material (ii) control rod performance and (iii) outcome of the examinations on the experimental irradiated sub assemblies..
The assessment of the mechanical properties of the highly irradiated fuel claddings during an RIA (Reactivity Initiated Accident) has been carried out in the framework of the PROMETRA programme. Three main types of tests including burst tests, hoop and axial tensile tests, have been performed in CEA-Saclay hot laboratories in order to determine the cladding tensile properties used in the SCANAIR code. The representativeness of each test with regard to the RIA loading conditions can be addressed and analyzed in terms of strain or stress ratio. The present paper reports the high strain rate ductile mechanical properties of irradiated ZIRLO TM and M5 TM alloys derived from the PROMETRA program and their comparison to the stress-relieved irradiated Zircaloy-4. Results of specific analysis of the behaviour of the 6 cycle M5 TM and ZIRLO TM 75 GWd/tM for temperatures higher than 600°C are also presented.
Journal of Nuclear Materials, 2008
This paper presents a unified phenomenological model to describe the anisotropic viscoplastic mechanical behavior of Cold-Worked Stress Relieved (CWSR) Zircaloy-4 fuel claddings submitted to Reactivity Initiated Accident (RIA) loading conditions. The model relies on a multiplicative viscoplastic formulation and reproduces strain hardening, strain rate sensitivity and plastic anisotropy of the material. It includes temperature, fluence and irradiation conditions dependences within RIA typical ranges. Model parameters have been tuned using axial tensile, hoop tensile and closed-end internal pressurization tests results essentially obtained from the PROMETRA program, dedicated to the study of zirconium alloys under RIA loading conditions. Once calibrated, the model provides a reliable description of the mechanical behavior of the fresh and irradiated (fluence up to 10 · 10 25 n.m −2 or burnup up to 64 GWd/tU) material within large temperature (from 20 • C up to 1100 • C) and strain rate ranges (from 3 · 10 −4 s −1 up to 5 s −1 ), representative of the RIA spectrum. Finally, the model is used for the finite element analysis of the hoop tensile tests performed within the PROMETRA program.
www-naweb.iaea.org
The Budapest Research Reactor is operated by the KFKI Atomic Energy Research Institute in Hungary. After full reconstruction it was started again in 1993, and a new irradiation rig called BAGIRA-1 (Budapest Advanced Gas-cooled Irradiation Rig Assembly) was installed in close cooperation with the Paul Scherer Institute (Switzerland). The rig is operational since 1998 and it is used to irradiate nuclear reactor vessel and fusion equipment materials in order to study and evaluate the irradiation ageing. The average fast neutron flux at the position of the rig is about 6x10 13 n/cm 2 s (>1MeV). The paper is a review of the material testing studies completed so far and the recent improvements of the rig.
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.
Preliminary work has been completed to evaluate the deformation behavior of the Generation II FeCrAl alloy class. The Generation II alloy, C35M, was neutron irradiated within the High Flux Isotope Reactor followed by post-irradiation examination to evaluate the mechanical properties, irradiated microstructure, deformation modes, and deformation microstructure. Complementary characterization techniques were used in this evaluation. It was found that after irradiation to 1.8 dpa at 364°C that the C35M specimen failed in a ductile manner. The observed deformation microstructure showed a high density of line dislocations, dislocation networks, and dislocation pileups. The observed performance and structures are consistent with literature studies on high-Cr FeCr alloys that were thermally aged at 500°C. The observed results suggest that Generation II FeCrAl alloys retain ductility under low-dose, prototypical LWR irradiation conditions.
Assessment of creep in reactor-irradiated CuCrZr alloy intended for the ITER first wall panels
Fusion Engineering and Design, 2018
CuCrZr alloy is candidate heat sink material for the ITER blanket, first wall, and divertor. During ITER operation it will be exposed to a combination of elevated temperatures, heat flux, and intense fast neutron radiation. This environment will challenge the performance of components and joints based on CuCrZr. To address this issue, mechanical tests were performed with irradiated and reference specimens of CuCrZr and its joints with 316 L(N)-IG (ITER Grade) stainless steel made by hot isostatic pressing. The reactor exposure up to ∼0.7 dpa was performed in the BR2 reactor at SCK•CEN, in water at a temperature of 257°C A special design was used to allow irradiation of specimens axially pre-stressed at different strain levels. The post-irradiation examination included: (i) tensile test, (ii) measurements of plastic deformation of samples axially loaded during irradiation (in situ creep test); (iii) thermal creep tests on irradiated samples. The fracture surfaces were examined in a hot cell using a Scanning Electron Microscope (SEM). The results were compared with data obtained from mechanical tests and SEM/EDX fracture surface analysis on non-irradiated reference samples. The level of a possible creep under irradiation is below the experimental uncertainty.
As part of studies conducted in France on Reactivity Initiated Accident (RIA), IRSN and EDF have launched a large experimental project (PROMETRA) carried out by CEA in order to provide both material properties and material failure data [1]. During the first phase of a RIA event, the in-service loading deforms the cladding in the circumferential direction under multiaxial tension, in a situation close to an axial plane strain situation. In order to accurately evaluate the risk of rod failure during this stage, it is important to develop models able to predict the material behaviour under those representative loading conditions. Obviously, the fracture behaviour has also to be determined. To this end, uniaxial tensile data have been obtained between 20°C and 1100°C under high strain rates (0.01 to 5s -1 ) and high heating rates (up to 200°C.s -1 ) from specimens machined along the axis of the cladding or in the circumferential direction (ring specimens).
Procedia Engineering, 2013
Development of materials plays a crucial role in the order to meet this objective, one of the methods is to largely limited by the void swelling and creep resist Prototype Fast Breeder Reactor (PFBR) is in advan PFBR with MOX fuel are alloy D9 as fuel clad components and piping and modified 9Cr-1Mo steel phosphorous and silicon contents in alloy D9 have modified version of alloy D9 as IFAC-1. Creep resis improved with the dispersion of nano-size yttria to de term creep strength, similar to D9, for increasing the 9Cr-1Mo steel wrapper for future metallic fuel reac Extensive studies on resistance of this new generatio material development. Improved versions of 316LN creep strength to increase the life of fast reactor and addition of boron to improve type IV cracking resist in ITER programme necessitates the development comprehensive research programme is being carri optimization of tungsten and tantalum contents for b mechanical tests including impact, tensile, creep and 2 wt. % and tantalum in the range 0.06-.014 wt., th found to possess better combination of strength and TBM is being manufactured by this RAFM steel. To programme has been initiated to develop advanced temperature of around 973 K and pressure of 300 b Inconel 617 superalloy tubes are indigenously develop