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S0022311519312292
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The hot deformation behavior of Zr 4 alloy at the deformation temperature range of 7501000C and the strain rate range of 0.00110 s
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The hot deformation behavior was investigated . The strain compensated constitutive model was established. The hot deformation activation energy was calculated. Murtys processing maps have been established . The microstructural observations were used to verify the Murtys processing maps .
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S0022311519312358
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Deposition of corrosion products on micro orifice specimens under accelerated flow conditions was investigated in high temperature water on several substrate materials and water chemistries 02ppm of Li 02.5ppm of H
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Presence of lithium reduces solubility of magnetite and CRUD build up rate. CRUD morphology is pH dependent particles dominated at neutral pH and crystalline at alkaline pH 2ppm of Li . The radial CRUD build up is independent of substrate material at neutral pH. The surface CRUD deposition depends on the electrical conductivity of the substrate material at alkaline pH. Electrokinetic activity drives the deposition of magnetite at alkaline pH.
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S0022311519312528
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Transmission electron microscopy was used to compare the microstructural defects produced in an Fe9Cr model alloy during exposure to neutrons protons or self ions . Samples from the same model alloy were irradiated using fission neutrons 2MeV Fe
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Proton irradiation redistributes alloyed elements without necessarily requiring a temperature shift to compensate dose rate. Vacancy dislocation loops were found to comprise the majority of damage after 2 MeV Fe irradiation. Proton irradiation lead to larger cavities indicating implanted hydrogen might play a role in their formation. phases were found at a density 9 higher when the dose rate from. was reduced by a factor of 3
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S002231151931253X
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This study focused on the preparation of metallic uranium and uranium zirconium alloys to measure the effect significant porosity had on thermal diffusivity from 20
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Samples of uranium metal and alloys of uranium 5 mass zirconium and 10 mass zirconium were prepared via powder metallurgy. By controlling sintering parameters samples with a porosity up to 70 were obtained. The thermal diffusivity of the samples was measured from 20C to 300C using the light flash technique. The thermal conductivity of the samples was calculated using literature correlations. Thermal conductivity as a function of porosity was plotted for each metal.
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S0022311519312668
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The fundamental understanding of interface properties is crucial in materials design and lifetime predictions . In this work the stability adhesion and impurity induced embrittlement of interfaces between tungsten and transition metal carbides have been investigated by first principles calculations . For all the systems the coherent W TMC interfaces show better stability with lower interface energies than the semi coherent W TMC ones . The impurities hydrogen helium oxygen and nitrogen tend to segregate to the coherent interfaces and act as strong embrittlers . Furthermore the interface could provide a low barrier channel to facilitate hydrogen and helium transport . The present work provides key mechanistic insights towards interpreting recent experimental studies of the interface structure and the hydrogen isotope retention in WZrC WTiC and WTaC materials under irradiation and guides the preparation of future W based materials with good resistance to irradiation damage .
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Coherent W 100 TMC 100 interfaces have better stability than the semi coherent W 110 TMC 100 ones. The impurities H He N O S and P tend to segregate to the interface and act as strong embrittlers. Interface provides a rapid channel to facilitate H and He transport along the interface. W based materials with a multi scale interface structure are suggested to enhance the overall performance.
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S0022311519312772
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In the present paper one proposes an equation of state allowing to reasonably describe the behavior of the xenon krypton mixtures in any proportion confined in the form of nanoscale bubbles in the UO
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An equation of state is proposed for confined Xe Kr mixtures in the UO. and MOX nuclear fuels. The parameters of the equation of state were determined using accurate molecular dynamics simulations results. The equation of state proposed here could be easily implemented in the nuclear fuel performance codes.
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S0022311519312991
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During the dissolution of the spent nuclear fuel some radionuclides are released to the solution simultaneously from different sources in the fuel . This is of particular importance to some radionuclides that contribute to the Instant Release Fraction which govern the initial radiation dose during the dissolution of the SNF . In this work a model that is able to discriminate between the different contributions responsible for the total concentration of a radionuclide in solution was developed . The model permits to establish that uranium and radionuclides that dissolved congruently with the UO
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Model developed to establish the contribution of segregated phases to SF dissolution. Mo Tc Pu Am Ce and La dissolve congruently with the matrix. Ru and Rh segregated from the matrix. Cs Rb Sr partially segregated to grain boundaries 2.1 0.9 and 0.6 respectively .
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S0022311519313078
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Large scale molecular dynamics calculations have been carried out to investigate the properties of nanometric helium bubbles in silicon carbide as a function of helium density and temperature . A dedicated interatomic potential has been developed to describe the interactions between helium and SiC atoms . The simulations revealed that the helium density can not exceed a certain threshold value which depends on temperature because of the plastic deformation of the SiC matrix . Both local amorphization at low temperatures and nucleation and propagation of dislocations at high temperatures have been identified as activated plasticity mechanisms . This work also predicts that very high pressure up to 60GPa could be reached in helium bubbles in silicon carbide .
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He bubble pressure in SiC is limited by plastic yielding. Plasticity mechanisms are amorphization and nucleation migration of dislocations. High He pressures up to 60GPa are possible in these bubbles.
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S0022311519313236
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Fracture toughness is one important mechanical property of reduced activation ferritic martensitic steels which are primary candidate structural materials applied to fusion reactors . Temperature effect on fracture toughness of the Chinese low activation ferritic martensitic steel was investigated in the range of 25550C with miniature three point bend specimens using the digital image correlation method to measure load line displacement . Results show that the fracture toughness J
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The fracture toughness J0.2 B of CLF 1 steel decreases from 25C to 450C and increases from 450C to 550C. The load line displacement of miniature 3PB specimens was measured by a 3D DIC method. Few typical dimples observed at 450C indicated quasi cleavage like features.
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S0022311519313583
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Carburization of structural alloys used in Fluoride Cooled High Temperature Reactors can result from the interaction between the metal and graphite components and cause changes in alloy properties . However carburization requires transport of carbon from graphite to alloy surfaces through the molten fluoride coolant . This work introduces an alternative to previously proposed carbon transport mechanisms . The new mechanism is based on the generation of carbonate ions CO
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This work introduces a carbon transport mechanism in molten fluoride salt systems based on CO. impurities. CO. acts as a salt soluble carbon bearing intermediate species that can be spontaneously reduced on metal surfaces. CO. can be generated by reaction between dissolved oxides and graphite.
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S0022311519313613
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CuCrZr alloy has been widely considered to be used as heat sink materials in the fusion reactor due to its superior properties . However CuCrZr alloy suffers significant creep deformation and softening as the temperatures is above 300400
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A novel Cu based alloy is designed and fabricated. High friction of low CSL grain boundaries is introduced into the matrix. High density Cr. Fe particles and Cr particles are detected in the matrix. The alloy exhibits high thermal stability even annealed at 500. C for 72h.
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S0022311519313844
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In the design of used nuclear fuel containers for deep geological repositories copper is considered to be a suitable and long lived barrier for corrosion resistance . The microstructures of state of the art copper materials used in this application produced through extrusion a grain boundary engineered electrodeposition technique and cold spraying were studied via electron backscattered diffraction . Desirable microstructural characteristics for localized corrosion resistance of pure copper were compiled from the literature considering grain size grain boundary character distribution and crystallographic texture . The subject copper materials were found to have favourable microstructures for localized corrosion resistance in particular a high fraction of special grain boundaries especially 3 twins rendering them suitable for the given application .
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Cu coating materials for used nuclear fuel containers were characterized with EBSD. Both electrodeposited and cold sprayed Cu contain many special grain boundaries. The microstructural features are favourable for localized corrosion resistance.
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S0022311519313868
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Bridging lower length scale calculations with the engineering scale simulations of fuel performance codes requires the development of dedicated intermediate scale codes . In this work we present SCIANTIX an open source 0D stand alone computer code designed to be included coupled as a module in existing fuel performance codes . The models currently available in SCIANTIX cover intra and inter granular inert gas behaviour in UO
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The characteristics of the SCIANTIX computer code are described. The models currently available in SCIANTIX are detailed with all the parameters . The verification of the numerical solvers is presented. Showcases of validation in both constant and transient conditions are presented. Future short and long term development plans are outlined.
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S0022311519313960
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Stable sodium zirconium phosphate iron phosphate glass ceramics for nuclear waste immobilization were synthesized by a melt quenching process and the effect of B
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NZP iron phosphate glass ceramics were synthesized by a melt quenching process. addition is the key point to achieve the NZP iron phosphate glass ceramics. Reasons on the formation of NZP due to B. addition were explained. Melt quenching process is a possible technology to prepare certain glass ceramics. NZP phosphate glass ceramics are potential matrix for disposal of specific HLW.
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S0022311519313972
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The influence of high flux helium plasma irradiation on compatibility between liquid lithium and CLF 1 steel a reduced activation ferritic martensitic steel was investigated since CLF 1 shows potential application as substrate material for liquid lithium plasma facing components . After exposure to helium plasma with flux of 510
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The compatibility between liquid lithium and high flux helium plasma irradiated CLF 1 steels is firstly reported. The fuzz induced by helium plasma irradiation improves the wettability of liquid lithium on the surface of CLF 1. The porous fuzz nanostructure significantly aggravates the lithium corrosion behavior of CLF 1 in liquid lithium.
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S002231151931400X
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A mechanistic model is developed for the force depth relationship of ion irradiated materials which is conducted by spherical nano indentation . With irradiation effect the pop in phenomenon almost disappears that is ascribed to the irradiation induced defects serving as dislocation nucleation sites that facilitate the generation of new dislocations . After materials yielding the evolution of statistically stored dislocations geometrically necessary dislocations and irradiation induced defects mutually contributes to the force depth relationships with irradiation effect . Thereinto the increase of loading force originates from the impediment of slipping dislocations by irradiation induced defects . By comparing with the experimental data of Fe12Cr alloy a reasonable agreement is achieved .
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A model is proposed for the force depth relationship of ion irradiated materials. Three distinguishing deformation stages for the force depth curve are characterized. Weakened pop in events and irradiation hardening are addressed. Theoretical results can match well with corresponding experimental data.
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S0022311519314011
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In this paper the effects of hydrogen content and temperature on the fatigue crack initiation and propagation behavior of Zircaloy 4 are investigated by
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Fatigue lifetime first increases and then decreases with hydrogen content. At RT fatigue crack growth rate increases with hydrogen content. Fatigue crack propagation path is affected by the hydride and grain boundary. Hydrogen content affects the fatigue crack initiation mechanism.
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S002231151931414X
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The presence of stable facets of nanobubbles in crystal lattice can significantly affect their diffusion coefficient but the existing theory of this phenomenon is too general and can not take into account atomistic structure of nanobubbles in a given material . Such a theory for the mechanisms of bubble motion in crystals can be extended and developed using methods of atomistic modelling . In this work we consider the movement of bubbles in the bcc lattice of U . The Beeres theory of faceted bubble motion is revised and a method of non equilibrium accelerated molecular dynamics in a pressure gradient is proposed . The results of the accelerated method calculations for U are verified using generic molecular dynamics calculations of free nanobubble diffusion . The new method significantly accelerates calculations of the diffusion coefficient for nanometer sized bubbles and opens the way for more accurate material specific calculations of gas filled nanobubbles diffusivity in nuclear fuels .
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A new molecular dynamics method is proposed for the calculation of faceted nanobubbles diffusivity in crystal lattice. Stable 110 facets significantly affect the diffusion rate. The activation mechanism of facets reconstruction is revealed. Results of the new method are cross checked by the modelling of free diffusion. Method can be used for other nuclear fuels.
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S0022311519314175
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The stress corrosion cracking behavior of 316L stainless steel forged to 0 12.6 21.4 and 39.8 reduction in thickness was investigated at constant K in light water reactor environments . The yield strength specimen orientation and water chemistry were correlated with crack growth rate and their dependencies are discussed . The crack growth rate of 316L SS increased monotonously with yield strength irrespective of the specimen orientation or water chemistry . Higher CGRs were observed when cracks propagate along the plane parallel to forging plane than normal to forging plane . The effect of local deformation on the anisotropic cracking behavior for different orientations crack paths and CGRs are also discussed .
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Yield strength is a strongly indicator to the SCC susceptibility of deformed 316L stainless steel. The anisotropic cracking behavior of tested samples arise from the different distribution of local deformation. The increased deformation in grain interior causes the change of cracking morphology.
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S0022311519314217
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The influence of implantation induced point defects on SiC oxidation is investigated via molecular dynamics simulations . PDs generally increase the oxidation rate of crystalline grains . Particularly accelerations caused by Si antisites and vacancies are comparable and followed by Si interstitials which are higher than those by C antisites and C interstitials . However in the grain boundary region defect contribution to oxidation is more complex with C antisites decelerating oxidation . The underlying reason is the formation of a C rich region along the oxygen diffusion pathway that blocks the access of O to Si and thus reduces the oxidation rate as compared to the oxidation along a GB without defects .
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All types of point defects accelerate oxidation of single crystal SiC. Defects can either accelerate or suppress the intergranular oxidation in SiC. C depletion at grain boundaries accelerates oxidation. C enrichment at grain boundaries suppresses oxidation.
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S0022311519314242
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The mechanism of irradiation creep and its effect on the internal stresses of reactor core graphite components are not well understood . Historical studies on Materials Test Reactor creep specimens have shown that thermal annealing can lead to recovery of creep strain . By selecting appropriate sets of trepanned graphite specimens for the AGR cores the thermal annealing programme in this work uses the findings of these historical studies in order to investigate irradiation creep and importantly the stress state of AGR core bricks . Changes in dimensions Coefficient of Thermal Expansion and Youngs modulus after each annealing step are reported . Laser Raman spectroscopy is used to investigate the effect of thermal annealing on irradiated graphite . The results are combined with core inspection data and demonstrate the complexity of estimating irradiation creep in different parts of reactor graphite components .
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Innovative approach to assess the stress state of graphite core components. Comparison with core inspection data. Unlike Materials Test Reactor specimens reactor specimens experience a multiaxial stress state. First Laser Raman study into the effects of thermal annealing on irradiated graphite.
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S0022311519314278
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Effects of chemical composition ion irradiation dose and temperature on unfaulting of irradiation induced Frank dislocation loops to perfect loops in two nickel based single phase solid solution alloys Ni20Fe and NiFe20Cr have been studied . The fraction of Frank loops decreases with irradiation dose from 7.2 to 38.4 dpa at 500
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High irradiation dose and temperature favor unfaulting of Frank dislocation loops to perfect loops that can reduce radiation hardening. The process of loop unfaulting may be hindered by sluggish diffusion in chemically more complex alloy.
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S0022311519314436
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Density functional theory calculations were performed to study reactions on uranium mononitride surface when it is in contact with environment gases such as O
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DFT calculations were performed on the adsorption and reaction behaviors of O. H. O N. and H. on the UN 001 surface. When T 350K the UN surface is totally occupied by oxygen within a short time scale from hours to 10. s. The diffusion of adsorbed O into the sub layer occurs only after full coverage of O on the UN surface.
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S0022311519314503
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Titanium is being considered as a candidate material for high level nuclear waste container due to its superior corrosion resistance . However problems may arise when titanium comes in contact with hydrogen bearing environments as it is vulnerable to hydrogen embrittlement . To assess the lifetime of titanium container hydrogen content and hydrogen permeation efficiency were estimated in infiltrated water through bentonite prepared with simulated groundwater of Beishan the preselected area in China by electrochemical and metallurgical analysis . It is concluded that the absorbed hydrogen due to general corrosion will not be sufficient for hydrogen embrittlement to occur in titanium for at least 10 000 years after deep geological disposal . This represents more than 300 and 5 000 half lives of Cs137 and Cs134 respectively cesium being the most likely element to escape a repository due to its high solubility in groundwaters .
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Hydrogen contents of titanium in deep geological disposal conditions were estimated. Effect of hydrogen distributions on hydrogen embrittlement HE was studied. The susceptibility of titanium container to HE was estimated over a disposal period of 10 000 years.
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S002231151931462X
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Under nuclear fusion environments displacement cascades in the potential first protective wall material vanadium and its alloys lead to a large number of point defects . As sink of point defects grain boundaries are observed to significantly affect the radiation resistance of structured metals . By using potential energy surface searching tools the interaction between the interstitial loaded 3 110 111 symmetric tilt grain boundary and the point defects in V based alloy is investigated . For the self interstitial atoms loaded STGB the vacancy located within certain distances from the STGB can be effectively annihilated within several nanoseconds via the interstitial emission mechanism . If an alloy doping interstitial as a Cr atom substitute the SIA or the vacancy is stabilized by alloy solutes as Ti atoms forming a solute vacancy complex near the STGB IE works more effectively with lower activation energy barriers and less thermal activation time . The results indicate that the STGB induced IE mechanism raises the radiation resistance of the V based alloy . The results contribute to the micro structure and constituent optimizations in designing an excellent V based first protective wall material .
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For the first time the self healing mechanism induced by grain boundary in an alloy is revealed. The self healing mechanism is confirmed to be the interstitial emission which recombines point defects at long time scale. The alloy studied is VCrTi ternary alloy which is expected to be a promising first protective wall material. The solute vacancy complexes promoted by non equilibrium segregation near a GB are focused and proved to facilitate IE. The uncovered differences between the effects of Cr and Ti on IE is beneficial to designing an V based alloy.
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S0022311519314631
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The release of fission product and salt compounds from a molten salt reactor fuel under accident conditions was investigated with coupled computer simulations . The thermodynamic modeling of the salt and fission product mixture was performed in The Gibbs Energy Minimization Software GEMS and the obtained compound vapor pressures were exchanged with the severe accident code MELCOR where the evaporation from a salt surface located at the bottom of a confinement building was simulated . The fuel salt considered in the simulations was LiF ThF
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Novel simulation of Cs and I release from the MSR fuel in accident conditions. Mixing effects in the salt reduce fission product release. Mixing affects the speciation of evaporated fission products.
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S0022311519314965
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Specimens of reactor vessel internals made of 18Cr10NiTi steel from the decommissioned VVER 440 unit after 45 years of operation were investigated . The aim was to assess the structure degradation degree responsible for the properties changes of RVI components . Radiation induced structural elements of RVI baffle former and flux thimble specimens with the damaging dose 7.943.0 dpa and irradiation temperature 280315C were studied using TEM SEM and APT methods . TEM studies showed the negligible swelling under these conditions with a slight increase of radiation swelling due to the formation of the low density large void population in case of higher temperature irradiation . APT studies revealed the formation of the high density NiSi rich clusters and an increase of its volume fraction with the temperature increase . The Orowan equation was used to assess the hardening contribution of different radiation induced structural elements . The observed yield strength increase is insignificant under the studied conditions .
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RVI samples from the decommissioned VVER 440 unit is studied by TEM SEM and APT. Void swelling observed in the studied dose range 843dpa is negligible. Higher irradiation temperature slightly increases the swelling by large voids formation. High density of NiSi clusters is observed presumably G and phase precursors .
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S0022311519315089
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In this study surface helium effect on hydrogen isotopes diffusion and trapping detrapping behavior was modelled and integrated into the HIDT simulation code . Effective dynamics properties of hydrogen in tungsten including diffusivity solubility recycling and diffusion barrier were considered to reflect the influence of He bubbles . Simulation results showed that total hydrogen retention was reduced with the existence of He bubbles near surface which was consistent with the reported laboratory experimental results . It was found that the most significant influence came from the diffusion barrier induced by He bubbles . With increasing the barrier factor total hydrogen retention changed from the tendency of decrease to increase . When the barrier factor was less than 0.3 hydrogen desorption from the implantation surface was dominant while that from the backside surface became dominant when the barrier factor was greater than 0.4 . In the meanwhile more hydrogen accumulated beyond the He bubble layer was observed . These hydrogen atoms occupied not only in trapping sites but also in lattice sites . Based on these findings three desorption stages namely surface desorption major desorption and backside desorption could characterize the TDS spectra with different mechanisms . In addition our findings were further substantiated by the reported experimental data . This study provides a new perspective to reveal the surface He effect on hydrogen isotopes retention behavior in plasma facing materials .
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Effective dynamics properties were considered to express the influence of He bubbles on hydrogen migration behavior. Total hydrogen retention was reduced with the existence of He bubbles near surface. Three desorption stages can characterize the TDS spectra with different mechanisms. Diffusion barrier induced by He bubbles showed significant influence on the desorption process. With increasing the barrier factor total hydrogen retention changed from the tendency of decrease to increase.
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S0022311519315284
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The liquid metal embrittlement sensitivity of the martensitic T91 steel by Pb Bi and LBE has been studied by small punch tests at different temperatures and strain rates . The LME occurrence in the three liquid metals depends on the strain rate and on the temperature . The T91 steel loaded in Bi Pb or LBE presents ductile behaviour at 300C and at 400C except in LBE at 300C at very slow strain rate for which fully brittle fracture surface was observed . It appears that the LME sensitivity of the T91 steel by LBE and bismuth is more important at 300C than 400C . No LME by lead has been observed . The most embrittling liquid metal is LBE then bismuth while the less one is lead . Some differences in the reactive wetting of the T91 steel by the saturated oxygen liquid metal could explain the difference in LME sensitivity in the three liquid metals .
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Study of liquid metal embrittlement sensitivity of the martensitic T91 steel by Pb Bi and LBE. The most embrittling liquid metal is LBE then bismuth while the less one is lead. T91 steel loaded in Bi Pb or LBE presents a ductile behaviour at 300C and at 400C except in LBE at very slow strain rate. According to the temperature and the strain rate observation of localized brittle fracture surface in presence of LBE and Bi.
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S0022311519315417
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The present work is concerned with the application of linear friction welding process accessible for the weld jointing of 9Cr reduced activation ferrite martensite steels . Optical and electron microscopic characterization of microstructures were performed in various regions of the weld joint . Hardness tensile and Charpy impact tests on the joined samples were processed to examine the mechanical property and reliability of the weld joints . It indicates that hot plastic deformation induced by linear friction triggers continual dynamic recrystallization in the weld zone along with high density dislocation substructures formed by such severe deformation which leads to a good combination of mechanical performances in the weld joint . Such linear friction welding comes beyond the rotational process restriction in conventional friction stir welding and avoid significant oxide inclusions porosities and coarsened grains brought by heat input as well . The work proves that the present LFW technique works well in the welding of 9Cr RAFM steels and inspires us of a future study on optimizing process parameters of the welding process for a better performance .
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Linear friction welding is applied in reduced activation ferrite martensite steels. Severe dynamic recrystallization refines the microstructure of weld joints. Complicated dislocation configurations enhance mechanical properties of weld joint. By optimizing the welding parameters in the study a better weldability is prospective.
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S0022311519315491
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Primary radiation damage featuring rapid atomic collisions and thermal spikes constitutes the foundation of a high fidelity description of radiation assisted microstructure evolution . To systematically describe the primary damage in the mixed fuel oxide systems
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A holistic functional form is developed to quantify the defect production. Defect size distribution is generalized with an exponential truncated power law. Vacancy clusters approach being charge balanced.
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S0022311519315508
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A batch of hot rolled pure W and doped W sheets under 27 90 rolling reduction were prepared aiming to elucidate how rolling reduction and impurity elements influence the microstructure evolution . EBSD results revealed the three stage evolution during hot rolling including original grain smashing recrystallization and fibrous process . The intermediate recrystallization leads to a typical bimodal morphology in pure W which facilitates the formation of fiber structure at a larger strain . It is the necklace shaped grains that cause non uniform stress and control the slip systems of 111 ND and 100 ND . The doped W sample exhibits a milder dynamic recrystallization behavior and a relative homogeneous recrystallized morphology due to the drag effect of impurities on grain boundaries that inhibit abnormal grain growth during dynamic recrystallization . It is believed that this homogeneous recrystallized morphology of doped W causes the discontinuity of fiber structure and aggravates the generation of the transversal cracks .
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Impurity induced microstructure change in hot rolled W is elucidated. The mechanism and diagram of microstructure evolution of hot rolled W is proposed. The bimodal morphology after DRX benefits the formation of fibrous grains. The cracks in hot rolled W is related to the discontinuity of texture.
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S0022311519315788
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Ferritic martensitic ODS steels are one of the candidate materials for Gen IV nuclear fission and fusion reactors . Residual ferrite was often found in the microstructure of 9Cr ODS steels . This constituent was reported to be responsible for the superior creep and high temperature strength . Using optical microscopy of an air cooled batch of ODS EUROFER inhomogeneous regions in the microstructure have been found with similar appearance to previously reported residual ferrite . In order to avoid a potential misinterpretation of inhomogeneous regions as residual ferrite detailed microstructural investigations have been carried out on the inhomogeneous regions using site specific nanoindentation scanning electron microscopy including electron backscatter diffraction and transmission electron microscopy . It is demonstrated that the inhomogeneous regions are free of oxide nanoparticles which possibly form due to imperfect mechanical alloying . These regions also exhibit lower hardness which is attributed to the absence of nanoparticles and a lower dislocation density . It is concluded that optical microscopy alone is insufficient to distinguish beneficial residual ferrite from undesired particle free regions . Our findings are underpinned by the consistency between the calculated theoretical yield strength the yield strength converted from the indentation hardness and the yield strength obtained from tensile testing .
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Inhomogeneous regions of ODS EUROFER were investigated using electron microscopy and site specific nanoindentation. Inhomogeneous regions are not residual ferrite as they exhibit lower hardness than the matrix and contain no nanoparticles. The absence of nanoparticles and a lower dislocation density are mainly responsible for the lower hardness of these regions. Optical microscopy alone is insufficient to identify inhomogeneous regions as being residual ferrite. Reasonable consistency in the yield strength obtained from different methods is achieved.
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S0022311519315909
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Multi metallic layered composite fuel cladding is designed to survive a longer period of time during accident scenarios for light water reactor fuel . One proposed MMLC cladding concept joining Zircaloy to steel via interaction barrier layers will increase corrosion resistance but result in a neutron absorption penalty . This cladding could also result in negative impacts due to mismatches between the mechanical behavior of its layers . In this study the mechanical performance of the MMLC has been evaluated in two ways small scale thermo mechanical simulations and full length fuel rod simulations . Both were carried out using the BISON fuel performance code . The small scale simulations predicted a reduction in residual stress and elastic deformation compared with standard Zircaloy cladding . The fuel rod simulations predicted decreased creep strain larger pellet clad gaps and possible plastic deformation under long time use when compared to Zircaloy and stainless steel cladding alone . The performance of various gap widths and layer thicknesses were compared to assist in the design of MMLC cladding we recommend an MMLC with a Zircaloy steel layer thickness ratio of at least 1.75 and an initial gap width of 40m .
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Simulations evaluated multi metallic layered composite MMLC cladding. No initial plastic deformation due to thermal stress was found. MMLC resulted in slow gap closure increased fuel temperature and increased fission gas release. We recommend a thickness ratio of at least 1.75 and a gap width of 40m
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S002231151931596X
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The void evolution in the presence of one dimensionally migrating self interstitials is considered . No saturation of the void swelling due to 1 D self interstitials is assumed . Void growth rate is considered to be driven by the dislocation bias and undergoes the stochastic fluctuations caused by the random nature of point defect jumps and collision cascade initiation . Void nucleation and subsequent growth are investigated . It is shown that initially nucleated random voids start shrinking already when the spatially aligned voids make a small fraction of the void ensemble . Despite a continuous generation of small void embryos in the random spatial positions and negligible vacancy emission development of the random void population becomes completely suppressed by the stochastic fluctuations when the aligned voids are the major sinks for point defects and the fraction of 1 D self interstitials is small but not negligible .
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No void swelling saturation. No vacancy emission. Growth of the spatially ordered voids at the expense of the random voids. Suppression of random voids growth.
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S0022311519315971
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The QUENCH LOCA bundle test series was launched to investigate the influence of the secondary hydriding phenomena on the applicability of the cladding embrittlement criteria . Seven out of pile bundle tests with different zirconium alloy based cladding materials were performed according to a temperature time scenario typical for a LBLOCA in German PWRs . Each bundle contained 21 electrically heated rods with length of about 2m . For two tests pre hydrided claddings were used . The profilometry measurements performed over whole length of the post test claddings showed formation of not only main ballooning area but also additional two or three ballooning regions . Cladding wall thinning from 725 to 350m due to ballooning was observed at the burst side along 50mm below and above burst opening . Oxide layer formed after the burst at the inner cladding surface around the burst opening with a thickness of about 15m decreasing to 3mat a distance of about 20mm from the burst opening . Hydrogen enrichments were observed for rods having been exposed to peak cladding temperatures of more than 1200K . The average maximal hydrogen concentration inside the hydrogen bands was less than 1000 wppm and the local absolute maximal hydrogen concentration in these regions was less than 1800 wppm . A part of the hydrogen absorbed inside the claddings formed the hydrides with m sizes which are distributed in the matrix intra as well inter granular . During quenching following the high temperature test stages no fragmentation of claddings was observed . Tensile tests performed at room temperature after bundle tests evidenced fracture at concentrated zirconium hydride regions for several rods with local hydrogen concentrations 1500 wppm and more . Claddings with lower hydrogen concentrations fractured due to stress concentration at burst opening edges . Other tensile tested claddings failed after necking far away from burst .
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Formation of several non symmetrical ballooning regions. Coolant channel blockage less than 35 . Burst temperatures between 1120 and 1140K depending on the cladding material. Burst temperature of pre hydrided claddings is 30K lower than for fresh claddings. Cladding fractures during tension at positions with hydrogen concentration 1500 wppm
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S0022311519316228
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We use atomistic simulations to investigate the interaction of vacancies and interstitials with interfaces between a crystalline metal and an amorphous covalently bonded solid . We select the gold silicon binary system as a model material and construct interface models along two different facets of crystalline Au and with amorphous Si created at three different quench rates . We compute formation energies of vacancies self interstitials and interstitial impurities as a function of position relative to the interface and find that they have markedly lower values near the interface than in the interior of the adjoining phases . We conclude that crystal amorphous metal covalent interfaces may be as effective at removing radiation induced point defects as interfaces in polycrystalline metals composites . Moreover irrespective of interface character the average formation energies of all point defects at all the Au a Si interfaces we investigated are comparable . Thus unlike in polycrystalline metals where an interfaces crystallographic character has a marked effect on its interactions with point defects all interface types in crystal amorphous metal covalent composites may be equally effective at absorbing all radiation induced defects .
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Atomistic models of crystalline gold Au amorphous silicon Si interfaces are developed. Formation energies of vacancies and interstitials are markedly lower near the interface than inside the adjoining phases. Irrespective of interface character the formation energies of all point defects at Au a Si interfaces are nearly identical. Interfaces between crystalline metals and amorphous covalently bonded solids are effective sinks for point defects.
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S0022311519316265
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Two narrow fractions of coal fly ash cenospheres with 90wt glass phase and different Si Al ratio shell porosity and thickness globule size and content of a mullite phase were used as a glassy source of Si and Al in the hydrothermal synthesis of pollucite analcime solid solutions at 150C and autogenous pressure 1.0M NaOH Cs Na ratios of 00.25 . Hollow microsphere globules of about 6070 and 140170m in sizes are produced as Cs immobilized forms in this process . The microspheres have a composite mullite ANA wall where ANA is analcime pollucite solid solutions . Composition and morphology of solid products as well as composition of ANA phases were characterized by PXRD SEM EDS and STA methods . The influence of the mullite content and shell thickness on the preservation of the product spherical shape has been established . The degree of Cs
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Microspheres with a composite mullite ANA wall are produced as a Cs mineral like form. Analcime pollucite phases with non significant difference in Si Al ratios were obtained. Mullite provides the preservation of a spherical shape of crystallized microspheres. The average Cs content in solid products is 2332wt .
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S0022311519316332
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This paper describes molecular dynamics simulations of fracture along a grain boundary in nickel decorated with helium filled bubbles . We find that the quantity governing embrittlement of the GB is its areal coverage by He bubbles . Indeed there is a threshold coverageapproximately 18 for the GB studied hereabove which the GB undergoes brittle fracture . During brittle fracture intergranular cracks advance by the growth and coalescence of He filled cavities ahead of the crack front . This process is enabled by plastic deformation of inter bubble ligaments . In models with lower He bubble coverage dislocation activity initiates in the adjoining grains prior to cavity growth and coalescence relaxing crack tip stresses and averting brittle crack propagation . We conclude that the ductile to brittle transition observed in this study is governed by a competition between dislocation emission ahead of the crack tip and localized deformation confined to the inter cavity ligaments .
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Molecular dynamics simulations of fracture along a grain boundary in nickel decorated with He bubbles is performed. When the fraction of boundary area covered by He bubbles exceeds approximately 18 the GB undergoes brittle fracture. During brittle fracture intergranular cracks advance by the growth and coalescence of He bubbles ahead of the crack front.
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S0022311519316526
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In this work numerical calculations have been employed to investigate the adsorption equilibrium state of hydrogen atoms on tungsten surface . The surface energy and relevant surface atomic density of different W surfaces are obtained through the empirical potential model . The calculations show that the higher surface atomic densities correspond to the more stable surfaces of which the most stable surface is the W surface . The H adsorption energies are systematically analyzed by density functional theory . The equilibrium concentration dependence for adsorption upon H coverage a temperature range from 300K to 1100K and a pressure range from 10
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Surface energies of various W surface are determined. Most of H desorb from the W 110 surface at 700K1100K and 10. Pa10. Pa. The monolayer H fluence on the W 110 surfaceat the equilibrium state is estimated.
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S0022311519316617
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Electron Backscatter Diffraction was used to investigate the grain boundary character and triple junction distributions as well as the microtexture on drawn pure and potassium doped tungsten wires . With an approximate diameter of 150m pure W wires were annealed at 1300 1600 and 1900C whereas K doped material was annealed at 1300 1600 and 2100C . The annealing was performed under hydrogen atmosphere for 30min . Both longitudinal and transversal sections were analyzed to assess anisotropic features . Up to 1600C all conditions presented a strong 110 fiber texture parallel to the drawing axis . With increasing annealing temperature the pure W material developed a more heterogeneous fiber texture while for the K doped material it remained homogeneous . Orientation correlation function analysis suggested sub grain coarsening as the recrystallization mechanism while grain boundary density and grain boundary character distribution exhibited anisotropic behavior as well as the triple junction distribution network . On the other hand the coincidence site lattices distribution did not present any anisotropy and followed the empirical law of the inverse cubic root of value . For all conditions the most abundant CSL boundaries were 3 9 11 17b 19a 27a and 33a . Based on the statistics of the triple junction types and their resistance to intergranular cracking it was revealed that increasing the annealing temperature might play a role in crack deflection since the resistance to intergranular crack growth is increased in the transversal section and reduced in the longitudinal section . This anisotropic behavior is preserved up to a higher annealing temperature in the K doped material .
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EBSD characterization is performed on pure and K doped W fibers. Strong 110 DA fiber texture is present regardless the annealing temperature. The most abundant low CSL boundaries are identified. Good agreement with the empirical law of the inverse cubic root of value was found. Anisotropy of triple junctions distribution is observed.
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S0022311519316691
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The FeCrAl Zy4 bi layer clad plate was fabricated by solid diffusion hot rolled and annealing treatment . The interface microstructure and phase analysis texture shear and tensile strength of the clad plate were investigated in details . The diffusional interface of clad plate was tightly bonded and there were no cracks and pores . The intermediate phases formed at the interface region of the plate were identified as ZrCr
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The intermediate phases formed at the interface were ZrCr. ZrFe. and Zr. Al. Zr. Fe formed at the interface of sample annealed at 600C for 3h and 5h. The textures of sample annealed at 600C for 5h are favorable for deformation. The optimal annealing process is 600C for 5h.
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S0022311519316721
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The mechanical properties of thin foils to be used as a targets window in a high intensity accelerator require non standard characterization techniques . In the current research the innovative small punch test technique had been used to map and determine the mechanical properties of SS 316L foil irradiated by high intensity of proton beams . The SPT results and the energy to fracture are presented and the fracture modes were determined by scanning electron microscopy observations and electron backscatter diffraction analysis . The irradiated samples were exposed to 3.6MeV proton bombardment at 250A and 290A for 42 and 5h respectively . The major damage as was reflected by significant ductility and energy to fracture losses was associated with the irradiated zones which experienced the highest temperature and protons flux . Based on the observed evidence of deformation twins and dense dislocation bands as well as the EBSD analysis it was revealed that the limited deformation in the irradiated samples is related mainly to radiation damage and probably a minor effect of hydrogen embrittlement phenomenon . The limited ductility was explained by the accumulation of radiation damage that most probably hinder dislocations mobility through crystallographic glide . Nevertheless these alternative deformation mechanisms involve nucleation and propagation of dislocation slip bands and intra grain fragmentation . The DSBs intersections were proposed as the source for stress localization which initiates crack formation which was followed by crack propagation through slip bands . This cracking mechanism was exhibited by a unique saw tooth fracture mode .
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SPT was proved as a highly sensitive tool to assess radiation damage RD effects. The proton damage mechanism is related only to radiation and not to trapped hydrogen embrittlement. A unique saw tooth fracture mode was found in the irradiated sample. Accumulation of proton damage accommodate alternative deformation mechanisms.
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S002231151931712X
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Chemical vapor deposited SiC and nuclear grade SiC SiC composite materials were exposed to light water reactor conditions at the MIT Nuclear Reactor . Three sets of samples were exposed within flowing PWR chemistry coolant under either 1 neutron and ionizing irradiation with resulting radiolysis products 2 radiolysis products without neutron flux 3 loop coolant . Post irradiation examination demonstrated an increase in corrosion from radiolysis products particularly H
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CVD SiC lost less than 1m in thickness year in BWR HWC conditions with 4wppm H. Radiation damage and radiolysis exacerbated corrosion significantly. Localized attack was observed only on samples exposed to damage and radiolysis.
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S0022311520300015
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The crack growth rates of un irradiated and un hydrided Zr 2.5Nb pressure tube material has been investigated in both air and in vacuum . Pre cracked CCT samples made from pressure tube were tested in air atmosphere at 300C and 350C at different stress intensity factors . It was observed that till a particular stress intensity factor there is no crack growth and after which the crack growth rate increases with increase in stress intensity factor . For the test carried out in vacuum it was observed that the crack growth rates for air is an order of magnitude higher than that of the crack growth obtained in the vacuum ambience .
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Stress assisted oxide cracking has been demonstrated for unhydrided unirradiated Zr 2.5Nb pressure tube material under static loading. The velocity of crack growth at 350C was determined as a function of stress intensity factor. SAOC was found to be associated with threshold stress intensity factor of 22MPam. The velocity of crack growth increased with increase in applied stress intensity factor. The SAOC is less likely event for propagation of initial flaws in PTs under operating condition of the PHWRs.
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S0022311520300088
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Pyrochemical reprocessing is being explored for reprocessing of spent metallic fuels of future fast breeder reactors . This process consists of several operations like electrorefining cathode processing and fuel casting process . The cathode processing and fuel pin casting processes are carried out at 10731723K . Hence the process vessels should withstand high temperatures molten metals and molten salts . Several structural materials and coatings have been proposed and evaluated for such applications . However high density graphite is the proposed material for process crucibles due to its high temperature compatibility . Bare HDG can not be used directly due to reactivity with molten uranium and its alloys . Hence coatings like yttria over HDG are being proposed to avoid undesirable carbon contamination of uranium . This review provides the various coatings and their performance evaluation for molten metal applications including our recent works on the development of thermal spray coatings for such an application .
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Structural materials for Pyrochemical reprocessing applications plays a crucial role. Graphite crucibles with Yttria coating is preferred for uranium melting application. SiC interlayer coating between graphite and Yttria provides better durability.
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S002231152030026X
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The sol gel route via internal gelation was applied for the production of Nd and Ce doped uranium dioxide microspheres . Trivalent and tetravalent Ce precursors were used and the influence of the precursors oxidation state on the fabrication process and the final product was studied . The successful introduction of the dopant into the 3UO
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Nd and Ce doped gels with. contents up to 30mol and good sphericity synthesised. successful implementation of the dopant into 3UO. 2NH. 4H. O matrix of the gel. Ce. more suitable than Ce. for the fabrication of doped UO. by IG. single phases for U. to U. Nd. and U. Ce. obtained. fabrication process for transmutation fuel or targets via IG improved.
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S0022311520300623
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The corrosion reaction of unirradiated uranium with deuterated liquid water under an argon overpressure was investigated . Two samples were examined at two temperatures under an argon atmosphere and contained conditions . The rate of corrosion was derived by monitoring the pressure changes in the cell as a function of time ascribed to D
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UD. was identified to form in the reaction products as part of the U D. Ar reaction. UD. formation confirms that hydride is produced from U and oxidation generated D. After a threshold pressure 0.5L is reached UD. formation is facilitated. Ar cover gas usage in enclosed systems could lead to pressure build up and further UH. formation.
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S0022311520300660
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Nuclear reactor lifetimes may be limited by nano scale Cu Mn Ni Si precipitates that form under neutron irradiation of pressure vessel steels resulting in hardening and ductile to brittle transition temperature increases . Physical models of embrittlement must be based on characterization of precipitation as a function of the combination of metallurgical and irradiation variables . Here we focus on rapid and convenient charged particle irradiations to both a compare to precipitates formed in NI and b use CPI to efficiently explore precipitation in steels with a very wide range of compositions . Atom probe tomography comparisons show NI and CPI for similar bulk steel solute contents yield nearly the same precipitate compositions albeit with some differences in their number density size and volume fraction dose dependence . However the overall precipitate evolutions are very similar . Advanced high Ni RPV steels with superior unirradiated properties were also investigated at high CPI dpa . For typical Mn contents MNSPs have Ni
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Nuclear reactor life times are limited by nano precipitates formed under irradiation. Ion irradiations provide a rapid affordable way to gain insight into precipitation. Atom Probe data estimates hardening and embrittlement at lower service dose. Advanced high Ni steels with superior unirradiated properties were investigated. Simple thermodynamic models are able to predict volume fraction for alloys.
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S0022311520300702
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Elucidating the synergistic effects of different energetic beams on the radiation response of nuclear materials is critical for developing an improved methodology for their evaluation when exposed to extreme environments . This article describes
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Sequential and simultaneous dual beam irradiations have been performed. on fine grained tungsten at high temperature. Loop damage is quantified as a function of dose or fluence and cavity damage is quantified at the end dose. Synergistic effects between sequential and simultaneous beams are revealed. Damage evolution is different in all cases.
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S002231152030074X
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In search for alternative cementitious materials for radioactive waste encapsulation geopolymers and inorganic polymers have received wide attention . Moreover Fe rich IPs offer an interesting alternative to high density concretes for use in radiation shielding applications . Materials can however be altered when subjected to ionizing radiation creating the necessity to evaluate the materials behaviour under irradiation conditions . In this study the effect of high dose rate gamma irradiation is investigated on CaO Fe
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Iron rich slag based IPs were irradiated with a 8.85kGy h. Co source. The curing time before irradiation affects the materials response. We observed strengthening of IPs associated to radiation induced iron oxidation.
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S002231152030129X
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of plasma facing components for DEMO divertor unravels new challenges to be met by the in vessel materials . Embrittlement induced by 14MeV neutrons in the baseline first wall material tungsten endangers structural integrity of PFCs . Chromium and or CrW alloy is currently considered as a candidate material in the design of mid heat flux PFCs as structural body of the monoblock . Cr has the superior mechanical properties in the low temperature range where the commercial tungsten products are brittle . However the fabrication of Cr requires high level purity control and is therefore challenging for mass production .
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Vacuum Arc Melter VAM is applied to produce pure Cr and Cr10W solid solution. VAM produced Cr exhibits the transition temperature for ductility being close to room temperature. Solid solution with 10 W significantly increases yield strength and work hardening at elevated temperature.
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S0022311520301392
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Zirconium alloy core components of nuclear reactors are manufactured following a series of cold working and heat treatment processes which alter the orientation of grains and hence their mechanical and fracture properties . Due to the hexagonally close packed nature of crystal lattice in these alloys the plastic deformation and the corresponding development of crack tip constraint is dependent upon the orientation of grains and the major texture . In this work a thin sheet of Zircaloy 4 has been used to fabricate deeply cracked compact tension specimens with initial cracks orientated along the rolling and transverse directions respectively . The fracture specimens have been tested at 25 and 300 deg . C which is the operating temperature of pressurized heavy water type reactors . The crack initiation toughness values have been measured through a modified blunting line expression as well as through measurement of width of stretched zone prior to stable crack growth from scanning electron microscope images . The values of crack initiation toughness as measured through the later method is lower compared to the former signifying physical initiation of crack unlike the approximation associated with 0.2mm offset line . In addition the initiation toughness as well as the fracture resistance for specimens with cracks oriented along the rolling direction of the strip are higher when compared to corresponding data for specimens with cracks oriented along the transverse direction . These differences have been explained from point of view of orientation of grains along the rolling direction during the fabrication process of these sheets . The grains plastically deform easily for loading along the transverse direction which is the mode I loading direction for cracks oriented along rolling direction . The associated loss of crack tip constraint raises the fracture resistance for rolling direction cracks .
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Specimens with initial cracks along transverse and rolling directions of Zircaloy 4 sheet are tested. Estimated crack initiation toughness using blunting line method is higher compared to that of SZW method. The initiation toughness and slopes of J R curves for rolling direction cracks are more. The effect of orientation of HCP grains on material plastic flow and fracture properties are opposite. The anisotropy in fracture properties can be explained on basis of orientations of the basal planes.
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S002231152030146X
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In a D T fusion reactor self sufficiency of tritium is one of the critical issues to maintain steady state operation of the reactor . Li
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Tritium release in neutron irradiated Li. TiO. and Li. TiO. Li. SiO. was obtained. The kinetic parameters of tritium diffusion in Li. TiO. have been obtained. One trapping site in neutron irradiated Li. TiO. but three in Li. TiO. Li. SiO
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S0022311520301525
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With the aim of improving the oxidation resistance of zirconium alloy claddings in light water reactors and enhancing the deposition efficiency and economics of surface coating electroplating was used to deposit a Cr coating with a rate exceeding 12m h on the surface of thermally treated Zr 4 alloy predeposited with an intermediate transition nickel layer . The evolution of helium bubbles in the cross section of the Cr coated Zr 4 alloy during 30keV He
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Cr coated Zr 4 alloy was successfully prepared by electroplating. Ni layer was designed to eliminate interface defects and enhance interface binding force. Metallurgical bonding was formed at the interface between each layer. The bubble evolution in Cr coated Zr 4 alloy irradiated by He. was. analyzed. A new. irradiation facility with TEM the only in running one in China was introduced.
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S0022311520301550
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The early stage thermal age hardening and the resultant nanoscale structure of 12Cr 12Cr7Al and 18Cr9Al ODS steels have been studied by Vickers hardness measurements and atom probe tomography respectively after isothermally ageing at 475C for 300h . The thermal age hardening of 12Cr 12Cr7Al and 18Cr9Al ODS steels was measured to be 4 HV 45 HV and 34 HV respectively . The results of APT analyses revealed that no phase separation occurred in all the three ODS steels while a significant precipitation of enriched phase with crystallography closely related to Heusler type Fe
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Early stage ageing behavior of FeCrAl ODS steels was studied at 475C for 300h. No phase separation was observed in 12Cr 12Cr7Al and 18Cr9Al ODS steels. Early stage hardening of 12Cr7Al 18Cr9Al ODS steels is due to phase formation. The precipitation of phases has no relationship with phase separation. The fast precipitation of phases is due to the high diffusivities of Al and Ti.
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S0022311520301586
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Hydrazine is added into secondary circuit in order to eliminate oxygen and to impose a reducing potential . Another possible role of hydrazine is to reduce hematite potentially present and thus reduce risk of stress corrosion cracking of steam generator tubes . However the latter effect of hydrazine is still not clear . In order to better understand the mechanism and the kinetics of hematite reduction by hydrazine under secondary circuit conditions an experimental study has been carried out . It was found that the reduction of hematite proceeds through the adsorption of hydrazine surface reduction and dissolution of hematite and coprecipitation of well crystallized octahedral magnetite .
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Ferric oxides are found in secondary circuit despite thermodynamic instability. Hematite can have a detrimental effect toward steam generator tubes. Hematite can be reduced by hydrazine in secondary circuit conditions. Slow reduction kinetics allows the persistence of hematite in industrial conditions.
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S0022311520301641
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Understanding the hydride platelet reorientation in zirconium alloy fuel claddings that potentially occurs during the dry storage process of spent nuclear fuel is of great technological importance for the cladding integrity . In this study detailed microstructural and crystallographic analysis of the circumferential and reoriented radial hydrides in a ZrSnNb cladding tube has been performed . The results show that the crystallographic nature of the reoriented radial hydrides in the material remains the same as that of the circumferential hydrides . The radial hydrides remain to be the
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An internal pressurization approach was proposed to apply the hoop stress on the Zr cladding tube. Detailed microstructural and crystallographic analysis on stress reorientation effect of hydrides was performed. The misfit strain accommodation mechanism was discussed by micromechanical stress analysis.
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S0022311520301951
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While degradation of concrete due to irradiation is a widely studied subject the deterioration and changes in microstructure of cement pastes immobilizing evaporator concentrates due to gamma radiation is a very specific subject because of the nature of the solidified radioactive waste . The cement paste proposed as possibly suitable binder for cementitious composites immobilizing evaporator concentrates a paste with the same composition but mixed with simulated evaporator concentrates and a reference paste were subjected to gamma irradiation from Co 60 source at doses of 2MGy and studied in terms of microstructure changes using Scanning Electron Microscopy with Energy Dispersive X ray Spectroscopy X ray diffraction and nanoindentation . SEM EDX revealed rise of new phases on the NP C samples rich in Na
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XRD revealed presence of charlesite instead of ettringite in composites immobilizing evaporator concentrates. Nanoindentation indicated formation of a new phases due to irradiation probably carbonates. Irradiation of composites with evaporator concentrates causes formation of sodium carbonates.
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S0022311520302051
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High purity SiC and SiC SiC composites coated with commercial TiN Cr CrN or CrN Cr multilayer coatings were irradiated in Ar or flowing PWR water in the Massachusetts Institute of Technology Nuclear Reactor Laboratory . Irradiation in Ar was performed in the core . In the water environment identical samples were placed in one of three different locations in core providing exposure to neutron damage and radiolysis affected water above core where samples were exposed to radiolysis affected water but not neutron damage or outside of the core where samples were exposed to the coolant water without the effects of radiation . Radiation in Ar revealed significant cracking of all but the TiN coatings attributed to differential swelling between the coating and substrate . Lattice swelling was not observed in any of the coatings but 0.2 void swelling was observed in the Cr coating . All of the coatings failed during water exposures in the core . CrN Cr spalled in each condition . Cr was protective except under radiation damage as a result of cracking and TiN severely degraded in the core with no coating was found following exposure . A SiC coating ATF cladding system is expected to perform adequately following improvements in coating ductility and purity .
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Four commercial coatings on SiC were irradiated in Ar and PWR water. Differential swelling led to failure of the Cr and CrN coatings. TiN was resilient to irradiation in Ar but failed under irradiation in water. Improvements in coating ductility and purity are expected to improve performance.
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S0022311520302087
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In this study the process involved in the fabrication of a potential accident tolerant fuel is described . Homogeneous uranium nitride microspheres doped with different thorium content were successfully manufactured using an internal gelation process followed by carbothermic reduction and nitridation . Elemental analysis of the materials showed low carbon and oxygen content the two major impurities found in the products of carbothermic reduction . Uranium nitride microspheres were pressed and sintered using spark plasma sintering to produce pellets with variable density . Final density can be tailored by choosing the sintering temperature pressure and time . Density values of 7798 of theoretical density were found . As expected higher temperatures and pressures resulted in a denser material . Furthermore a direct correlation between the onset sintering temperature and thorium content in the materials was observed . The change of onset temperature has been related to an increment in the activation energy for self diffusion due to the substitution of uranium atoms by thorium in the crystal structure .
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Homogeneous thorium nitride and uranium nitride solid solution was achieved. Densities exceeding 90 of theoretical density were measured in produced pellets. Influence of thorium content on sintering onset temperatures during SPS was observed. No blackberry structure was retained in the produced pellets using SPS. Low porosity observed at temperatures of 1550C and above.
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S0022311520302269
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Deuterated polymer beads with specifications on their size and structure are needed for preparing the targets in the inertial confined fusion experiments . Driven by the need to fabricate these beads meeting physical design the coflow microfluidic device composed of glass round capillaries were fabricated . The formation mechanism of the droplets of the polymer solution the vacuole formation in the polymer beads and the size relationship between the O droplets and the resulting polymer beads were also investigated . The diameter of the O droplets can be controlled by designing microfluidic device with appropriate capillaries and tuning the flow rate of the polymer solution and aqueous solution . Residual solvent in polymer beads was reduced by optimizing the solidifying and drying process leading to the decrease of the vacuoles . The size relationship between the initial O droplets and the resulting beads was revealed and confirmed . Consequently monodisperse polymer beads such as 480m and 700m poly 700m polystyrene and 800m deuterated polystyrene beads meeting physical design were successfully prepared which can be used to prepare the target in the inertial confined fusion experiments . It demonstrated the technical feasibility of controlling the bead size providing a promising approach to fabricate the polymer beads meeting size requirement .
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Monodisperse polymer beads meeting physical design of the targets were successfully prepared. Diameter of the droplets can be controlled by designing microfluidic device and tuning the flow rate of the phases. Residual solvent in polymer beads was reduced suppressing the vacuole formation. Size relationship between the initial O droplets and the resulting beads was revealed and confirmed.
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S0022311520302294
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The effect of temperature on the liquid metal embrittlement susceptibility of an Fe10Cr4Al ferritic alloy exposed to stagnant oxygen depleted lead bismuth eutectic was investigated using slow strain rate tensile tests . The results show that the total elongation to rupture of the specimens tested in LBE decreases with increasing temperature from 150 to 350C while an upward trend is observed in the temperature range of 350500C . Such a strong temperature dependence is called a ductility trough . This phenomenon is not observed in the specimens tested in Ar reference environment . Fractographic examinations reveal that LME occurs at 150 up to 450C and ductility fully recovers only at 500C . In addition transgranular cleavage and intergranular cracking are the major fracture modes on the fracture surfaces affected by LME .
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A ductility trough is found and the alloy is embrittled at 150450C. Ductility recovers only at 500C. Cleavage and intergranular cracking are the major LME failure modes.
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S0022311520303135
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Thermal ageing behaviors of 15wt . Cr oxide dispersion strengthened steels with the contents of Al of 5 7 and 9wt . which were isothermally aged at 475C for 9000h have been investigated by Vickers hardness measurement tensile test and atom probe tomography to correlate the age hardening with the evolution of nanometer scale structures for understanding the age hardening mechanisms . The age hardening of 15Cr ODS steel is attributed to phase separation . The age hardening of 15Cr5Al ODS steel is considerably higher relative to 15Cr 15Cr7Al and 15Cr9Al ODS steels which is due to the strengthening by both Cr enriched phases and enriched phases formed in 15Cr5Al ODS steel . In 15Cr7Al and 15Cr9Al ODS steels however the age hardening is only attributed to the phase strengthening because phase separation is completely suppressed in the ODS steels when the content of Al is
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Long term ageing behavior in 15Cr ODS steels with 5 7 9wt Al was studied. 15Cr5Al ODS steel exhibits the strongest age hardening due to and phases. phase separation is completely suppressed in 15Cr ODS when Al content. 7wt . Age hardening from phase strengthening gets stronger with Al content increasing. Age hardening mechanisms are proposed for the three Al alloyed 15Cr ODS steels.
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S0022311520303160
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We report here on the generation of simulated defects on sintered uranium dioxide fuel pellets by taking advantage of the underwater laser ablation procedure and their subsequent characterization . This work we believe can play a role towards the validation of the performance of fuel pellet inspection machines . A repetitive fiber laser capable of delivering pulses of nanosecond duration in conjunction with a galvo scanner served as the machining tool in the experiment . The study of the dependence of mass ablation rate on laser fluence water column height repetition rate and beam scanning speed formed the bulk of this work . A water column of height 3mm above the pellet surface in combination with a laser fluence of lying within 67J cm
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Simulated defects were generated on sintered UO. pellets using a pulsed fiber laser. The defects e.g. pits cracks missing surface area were formed by laser ablation. The dimension of the fabricated defects can be controlled. Underwater ablation resulted in cleaner defects without any bulging near the edge. Defective pellets can be utilized for authentication of pellet inspection machines.
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S0022311520303251
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The Small Punch Test was developed as an alternative test to evaluate the mechanical properties of materials from a small material volume . Although the SPT is close to its standardization as a mechanical testing method there are many methodologies for the estimation of the different mechanical properties . All of them are based on the use of a correlation equation to link the SPT and the estimated mechanical property . The scattering of these correlation methods is generally great enough to make it necessary to delve deeper into the understanding of the physical behavior of the SPT .
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An optimized correlation method has been designed for the Small Punch Test SPT . Yield strength is obtained with this method reducing the experimental scattering. The optimized correlation method is an improvement of the t 10 offset method.
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S0022311520303330
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Environmentally assisted fatigue tests in borated and lithiated high temperature water were conducted to investigate the effect of dissolved oxygen strain rate and strain amplitude on fatigue life of 308L weld metal . It was found that fatigue lives were comparable in 0.005ppm DO and 0.1ppm DO water and decreased slightly with decreasing the strain rate from 0.04 s to 0.004 s. The EAF effects were more pronounced at low strain amplitudes . The EAF cracking mechanisms involved with dendrite boundary orientation relative to cyclic stress axis and ferrite in 308L weld metal are discussed .
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Low cycle fatigue life of 308L weld metal decreases in high temperature water. Dendrite boundary DB orientation to loading axis affects EAF cracking process. ferrites in 308L weld metal mainly inhibit crack growth in EAF cracking process. The EAF mechanisms involved with DB orientation and ferrite are proposed.
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S0022311520303810
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This paper presents the results of the complex structural studies of RPV steels from the decommissioned in 2017 VVER 440 reactor after 45 years of operation . During its operation the RPV beltline weld 4 had been annealed twice by mode 430C 150h in 1987 and after three fuel campaigns by mode 475C 100h . APT and TEM studies were used for characterization of the radiation induced precipitates and radiation defects and AES and SEM fractography for assessment of the grain boundary segregation level . Complex structural studies of RPV steel after such a long term operation are carried out for the first time . It is shown that the thermal exposure at 290C for 45 years leads to the formation of the Cu rich precipitates with a low number density . Recovery annealing results in a high level of P grain boundary segregation together with incomplete Cu return to the matrix . Re irradiation leads to further Cu and P matrix depletion through the formation of CuP radiation induced precipitates on the one hand and the P grain boundary segregation on the other . Eventually the saturation of secondary precipitate formation and deceleration of the P grain boundary segregation occurs at high fluence .
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RPV steels from decommissioned VVER 440 are studied by APT TEM AES and fractography. Low density Cu precipitation under thermal exposure 290C 45 years is observed. P grain boundary segregation in WM after annealing and re irradiation is observed. Under re irradiation CuP precipitation in WM saturates with the fluence increase.
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S0022311520303913
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The onset of breakaway irradiation growth in Zr and its alloys has been correlated with the nucleation and growth of faulted vacancy
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Breakaway irradiation growth is correlated with faulted basal c loop nucleation. Solute segregation is demonstrated to reduce stacking fault energies. Fe and Sn segregation reduces thermodynamic barrier for c loop growth. Sn only segregates to growing faulted c loop loops and not c loop precursors. Fe and Cr segregation can stabilize c loop precursors in the nucleation regime.
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S0022311520304013
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The evolution of the crystal lattice of samples made of UO2 doped with different concentrations of Nd in stoichiometric and hypo stoichiometric conditions has been systematically studied by X ray diffraction and X ray absorption spectroscopy . The substitution of a trivalent cation for the U
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Description of the solubility of the U. Nd. system at room temperature. Characterisation of the charge compensation mechanisms and local disorder for the U. Nd. system. Analysis of the crystal lattice of the U Nd O system assessed complementary by XRD and XAS.
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S0022311520304153
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The X501 transmutation experiment irradiated in Experimental Breeder Reactor II was intended to evaluate the safety and performance of the addition of minor actinides into the metallic nuclear fuel system .
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The X501 experiment demonstrated potential alloys for minor actinide transmutation. Performance of the X501 G591 metallic fuel pin is in line with historical experience from EBR II driver fuel. Fuel cladding chemical interaction FCCI characterization revealed Am infiltration into the cladding.
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S0022311520304554
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Precipitation hardening by precipitates is an effective strengthening mechanism for designing novel high entropy alloys . The stability of these precipitates under elevated temperature irradiation is a major concern for their application in nuclear industry . In the present study a strengthened FeCoNiCrTi
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A strengthened FeCoNiCrTi. HEA is irradiated with 6.4MeV Fe. ions at 400C 500 C and 600 C. Disordering and dissolution of precipitates are characterized by transmission electron microscopy and atom probe tomography. The ordered L1. structure of precipitate is susceptible to the radiation damage. The behavior of precipitates under irradiation is believed to be controlled by displacement damage and radiation enhanced diffusion.
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S0022311520304773
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Determining tensile properties from small punch test is being pursued actively in the nuclear industry due to the limited volume of material such tests use compared with standard tests which can be critical when considering active or development samples . One of the crucial challenges in harnessing the full potential of this technique is formulating methodologies which correlate the small punch specimens deflection to equivalent uniaxial tensile properties . Existing approaches for correlation rely on deflection obtained from a single point on the small punch test specimen used with empirical equations to make the correlation . However the deflection and strain accumulation in a small punch specimen is highly heterogeneous and data from a single point does not represent the gross deformation evolving in the specimen . This data when used in conjunction with the empirical formulations for deriving equivalent uniaxial tensile properties would not result in accurate identification of material properties . In this work we offer an alternative approach which uses the full field deflection of the specimen mapped through in situ digital image correlation . The use of digital image correlation combined with inverse finite element analysis augments the existing method of material properties identification from single point deflection data thereby significantly improving the reliability of the measurements .
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Full field deflection of small punch test specimen was mapped using in situ DIC. Deflections were used in inverse FEM for estimating mechanical properties. Better estimate of properties was achieved using deflections from multiple points.
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S0022311520305183
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Zircon is important ceramic used widely due to its excellent properties in particular for nuclear waste immobilization . However it is difficult to obtain mono phase zircon with high yield usually result in double phase ZrO
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Ce doped double phase 0.2Zr. Ce. Zr. Ce. SiO. 0. 1 ceramics were fabricated. Effect of Ce content on the phase and microstructure evolution was elucidated. ZrSiO. retained tetragonal while ZrO. transformed from monoclinic to cubic. Lattice parameters increased with Ce doping revealing lattice immobilization of Ce. Grain size and compactness increased with increasing Ce content.
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S0022311520305286
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A feasibility study of the cold spray deposition process of 304L stainless steel on 304 stainless steel substrates as a mitigation method for chloride induced stress corrosion cracking has been investigated under various substrate conditions . The study is aimed at the application of this technology to mitigate CISCC that may potentially occur in or near the fusion welded regions of stainless steel canisters in Dry Cask Storage System for used nuclear fuels . Spherical gas atomized 304L stainless steel powder in the size range of 25m44m was used as feedstock powder for the cold spray process . The powder was deposited on 304 substrates with various surface conditions as polished oxidized cold rolled and plates with prototypical CISCC . The effects of cold spray parameters on quality of cold spray coatings were investigated . Thickness porosity and phases in the as deposited materials were evaluated using scanning electron microscopy and X ray diffraction and correlated with microhardness and adhesion strength measured via micro indentation and ASTM C633 pulling test respectively . XRD analysis of the coatings was also conducted to examine the effects of cold spray condition on residual stress state in the coating . Detailed cross sectional examination of coating substrate interfaces was performed with transmission electron microscopy equipped with energy dispersive spectroscopy . Dense and continuous coatings with good adhesion strength high hardness and high degree of compressive stress were produced for the various substrate conditions by adjusting cold spray parameters . The results demonstrate that a cold sprayed stainless steel coating is a viable option to provide a physical barrier against CISCC in fusion weld regions of stainless steel in corrosive chloride salt bearing environments .
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Cold spray stainless steel coatings were developed as a CISCC mitigation approach for used fuel canister systems. Beneficial compressive residual stresses and high adhesion cohesion strength of cold spray coatings were achieved. Self cleaning effect during cold spray process disrupted oxide scale on the substrate enabling excellent adhesion. High quality stainless steel coatings were produced on sensitized substrates with prototypical CISCCs.
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S002231152030533X
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Embrittlement caused by tritium and helium produced as the beta decay product of tritium affects the structural integrity of materials used in nuclear fusion reactors . We assess the embrittlement of 316L stainless steel used as the primary vessel in the tritium storage and delivery system with a planned operation scenario for the International Thermonuclear Experimental Reactor . First based on the current system design and planned operation scenario the tritium and helium concentrations in the primary vessel material are calculated . The effects of tritium and helium on the mechanical properties are estimated in comparison with available experimental data . We conclude that if a forged material is used the mechanical properties of the primary vessel will not significantly degrade over planned operation in terms of fracture toughness . In addition fatigue and creep are expected not to be significant . Last we discuss two possible methods to lower the concentration of helium for the purpose of reducing the residual risk .
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Tritium and helium produced by tritium beta decay cause embrittlement in steels. Embrittlement of 316L SS used in T storage delivery system for ITER operation plan is assessed. If a forged material is used structural integrity will be maintained. Two methods are discussed to further reduce the risk of embrittlement.
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S0022311520305341
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Silicon carbide fibre in silicon carbide matrix composites are a promising cladding for use in accident tolerant fuels in current light water reactor designs . However as they are a radically different material from current metal clads current thermomechanical simulation methods struggle to accurately predict their behaviour especially regarding the potential development of cracks . Thus a new peridynamic model for SiC SiC cladding has been developed in the Abaqus finite element code . The material model was isotropic and considers matrix cracking and fibre pull out . The thermal expansion swelling and the degradation of the thermal conductivity are modelled under typical LWR irradiation conditions . The swelling on the outer surface is predicted to be greater than the inner surface due to the lower irradiation temperature causing a tensile stress on the inside of the cladding tension being more challenging for a ceramic than a metal . This stress increases during the decrease in power at the start of a typical pressurised water reactor refuelling outage and causes microcracking of the matrix on the cladding inner surface . In models without fibres cracks would propagate through the cladding . If fibres are modelled matrix cracking will extend to a depth of around 20 through the cladding from the inner surface which is unlikely to be an acceptable design . If an inner monolith of SiC is additionally modelled cracking propagates through the monolith and acts as a stress raiser for matrix cracking in the composite and therefore does not constitute a design improvement . If an outer SiC monolith is modelled fibre pull out strain on the inner surface of the cladding was increased by just under 70 . No cracks are predicted in an outer monolith which may therefore remain gas tight and thus a more suitable design . These predictions are consistent with experimental findings .
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Irradiation of accident tolerant SiC SiC cladding is modelled using peridynamics. Due to swelling a tensile stress is predicted on the inside of the cladding. In the absence of a monolith micro cracking extends to 20 from the inner surface. Cracks in an inner monolith are a stress raiser in the composite. No cracks are predicted in an outer monolith hence its the most promising design.
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S0022311520305353
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The hydrothermal corrosion of polished and as cut high purity chemical vapor deposited SiC was studied in a constantly refreshing water loop . Light water reactor conditions were simulated at 288 320 and 350C with dissolved gas concentrations between 0.15 and 3ppm H
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High purity SiC was exposed to high purity water with and without oxygen. Without oxygen mass loss was small and no localized attack observed. With oxygen mass loss was significant with extensive grain boundary attack. The reaction in O. proceeds through a single step with a reaction order of 1. An equation is given to predict the mass loss rate in hydrothermal conditions.
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S0022311520305663
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To develop advanced reduced activation ferritic martensitic steels for fusion reactor structural applications both carbonitride and carbide strengthened castable nanostructured alloys were explored for higher densities of MX nanoprecipitates . Systematic comparisons between the two types of CNAs indicated generally similar microstructures and comparable tensile properties and creep resistance . However the carbide CNAs did show some advantages over the carbonitride CNAs in terms of the uniformly distributed higher density of MC nanoprecipitates greater Charpy impact upper shelf energies less deuterium retention and swelling and potentially less transmutation induced composition changes and consequently thermodynamically more stable carbides . The carbide CNAs showed the best balanced high performance in the examined properties in contrast to the significantly varied performance of oxide dispersion strengthened alloys and the generally lower performance of current RAFM steels .
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Carbides provide better examined performance than carbonitrides to steels. Compared to carbonitrides carbides are more uniformly distributed in the matrix. Lower carbon and nitrogen content tend to yield greater Charpy absorbed energy.
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S0022311520306425
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Uranium carbides have attracted renewed interest as advanced nuclear fuels for Generation IV reactors . As an important property required for gas bubble modeling in nuclear fuels the surface energy of uranium carbides is scarce in literature . In this work we study the surface properties of uranium carbides by first principles density functional theory calculations . Surface orientations with maximum Miller index up to 3 2 and 2 are investigated for UC U
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Surface energies of uranium carbides predicted by first principles calculations. Surface morphologies of uranium carbide crystals obtained by Wulff construction. Surface stability analyzed based on surface termination and chemical potential.
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S0022311520307091
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U 6Nb is a uranium alloy containing 6wt niobium that possesses high corrosion resistance . The structure and composition of the passivating oxide layer formed on air aged U 6Nb which gives the material its corrosion resistant properties was characterized using surface scattering techniques . Stable oxide layers formed on the surface of a set of U 6Nb alloy thin films under ambient conditions were investigated using neutron reflectometry x ray reflectometry and grazing incidence x ray diffraction . The passivating oxide was composed of approximately 27 U 5 Nb and 68 O primarily consisting of a thin niobium oxide layer separating a thicker UO
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Passivating oxide on U 6Nb uranium niobium alloy films has a bilayer structure. A thin niobium oxide layer separates UO. layer from the underlying U 6Nb alloy. Niobium oxide enrichment at the alloy interface enhances corrosion resistance. Oxide formation doesnt significantly alter composition of the underlying metal. Neutron and x ray reflectometry provide surface oxide compositional depth profile.
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S0022311520307911
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Scanning tunneling microscopy has been performed on an electro polished gallium stabilized delta phase plutonium sample under ultra high vacuum conditions . The images obtained were used to measure surface roughness as ion sputtering and annealing was applied to the plutonium sample . It was demonstrated that sputter anneal cycles progressively increased the surface roughness of the plutonium . Additionally Auger and scanning tunneling spectroscopy measurements revealed an initially unstable and dynamically changing surface that became increasingly stable as sputter anneal cycles were applied . The stabilized surface region contained within the sputter crater was observed to be unchanged for up to 17 months after the final sputter anneal cycle .
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Ion sputtering increases roughness of plutonium surface. Scanning tunneling spectroscopy reveals dynamic surface changes. Consecutive sputtering and annealing stabilizes plutonium surface.
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S0022311520309922
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This paper presents the first comparative experimental assessment of the microstructures of the same surveillance specimens of VVER 1000 high Ni weld after the primary irradiation recovery annealing and re irradiation in the same industrial NPP to assess the possibility of its lifetime extension after annealing up to 60 years . Therein studies of both the initial and radiation induced hardening phases and the behavior of grain boundary segregation as well are presented .
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VVER 1000 high Ni weld is studied by APT TEM AES in primary and re irradiated states. parameters of hardening elements are equal in primary and re irradiated states. Grain boundary segregation level is the same in primary and re irradiated states. Re embrittlement rate of high Ni VVER 1000 welds should not exceed the primary one.
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S0022311520309958
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Lanthanide fission products such as neodymium formed during the irradiation of metallic fuels are known to cause deleterious effects from chemical interactions occurring at the fuel cladding interface a phenomenon known as fuel cladding chemical interaction . The use of fuel based additives that bind with the lanthanide elements within the fuel meat alleviating their interactions at the fuel cladding interface is one potential method proposed to mitigate the FCCI phenomenon and extend the burnup potential of such metallic fuel systems . In this study antimony is evaluated as one such additive and neodymium is used to represent the lanthanides . A Sb Nd alloy is fabricated which consists of two intermetallic phases SbNd and Sb
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Antimony is investigated as a potential fuel additive to mitigate fuel cladding chemical interactions FCCI . Neodymium used as one of the main lanthanide fission products causing FCCI. Diffusion couple experiments Nd HT9 led to the formation of Fe. Nd. and Fe. Nd. Diffusion couple experiments Sb Nd HT9 exhibited no adverse interactions. Density functional theory calculations substantiated a lack of chemical interactions between Fe and Sb Nd.
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S0022311520309971
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The key issue in tokamak is change of physical properties of plasma facing materials such as tungsten due to extreme conditions and long pulse operation . Practically it is difficult to analyze hardness changes online through conventional methods . Hence an online monitoring technique is much desired which measures hardness changes in fusion devices . In this study laser induced breakdown spectroscopy is used as in situ monitoring tool to measure hardness of tungsten heavy alloy samples after exposure to different irradiation of plasma ranging from 0.108 to 1.00MW m
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Exposure of tungsten heavy alloy with PSI plasma irradiation. Variation in hardness crystalline and microstructural properties of WHA with PSI power density. In situ LIBS hardness measurement of tungsten heavy alloy WHA . Relationship between LIBS calibration curve ionic atomic line ratio and plasma electron temperature with harness.
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S0022311520310047
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The effects of alternating dissolved oxygen and dissolved hydrogen on the corrosion behavior of Alloy 52 are studied by electric electrochemical methods measured in situ including contact electric resistance and electrochemical impedance measurements together with multiple ex situ surface analysis techniques including scanning electron microscopy and X ray photoelectron spectroscopy analysis . The film resistance changes cyclically in both DO and DH cycles . In DO cycle film resistance decreases with the increase of corrosion potential E
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Film resistance in DO cycle and DH cycle shows different trends with E. Pre immersion in high DH condition can enhance oxidation in DO condition. Immersion history in low DH condition is favorable for corrosion resistance.
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S0022311520310060
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An approach to improve the performance of steels for fusion reactors is to reinforce them with oxide nanoparticles . These can hinder dislocation and grain boundary movement and trap radiation induced defects thus increasing creep and radiation damage resistance . The present work investigates the thermal stability of the microstructure and the evolution of defects in a 0.3 Y
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Mechanical alloying introduces a deformation state that persists in ODS steels even during annealing at 1200 K. Deformation state and Y O based nanoparticles are likely responsible for microstructural instability at 1400 K. Positron annihilation results indicate that Y O based nanoparticles are able to trap thermal vacancy clusters.
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S0022311520310084
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Two flux thimble tubes made of 15 cold worked 316 stainless steel were harvested from Ringhals Pressurized Water Reactor Unit 2 with peak damages of 76 and 100 displacements per atom after 29 and 34 years service respectively . Specimens sectioned from parent tubes were comprehensively characterized with nominal damage levels of 0 41 74 76 and 100 dpa at a nominal temperature range of 285323C . Both FTTs contained helium and hydrogen gases as transmutation products . The helium follows a production rate of 9.8 appm dpa while environmental factors complicate hydrogen production obscuring an exact H dpa ratio . Irradiation induced dislocation loops nano cavities solute clusters and microsegregation were all observed . The dislocation loops and nano cavities indicated saturation at 41 dpa . The solute clusters continued to evolve with NiSi clusters formed at 41 dpa and NiSiMnP clusters formed at 74 and 100 dpa but neither clusters exhibited distinct diffraction patterns at any damage levels . Solute clusters were observed to frequently be co located with dislocation loops but fully decorated loops were rarely detected . Significant radiation induced segregation was observed around grain boundaries at all damage levels . The modified inverse Kirkendall model captured the RIS behavior of major elements . Large cavities within or around an MnS rich region were observed for the first time . Through all the damage levels void swelling is always below 0.05 making significant dimensional change unlikely in core internals when used at similar conditions . Meanwhile the role of overwhelming nanocavities presumably helium bubbles should be considered in other potential degradation mechanisms including irradiation assisted stress corrosion cracking embrittlement and loss of fracture toughness which remain the concerns for extended operation of nuclear power plants .
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Flux thimble tube made of commercial 316 SS was characterized after 34 years of service in a commercial nuclear power plant. The 316 SS had a recorded maximum damage level of 100 dpa. Significant void swelling is unlikely to occur in core internals made of 316 SS while operating at similar conditions.
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S0022311520310138
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Effects of cooling rates on mechanical behavior of steam oxidized Zircaloy 4 are systematically investigated with the cladding material oxidized to the current ductility based regulatory limits . Three different cooling rates associated with distinctively different quenching environments are tested ambient air cooled 7
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LOCA with different cooling rates and Ring Compression Tests RCT conducted. Pre hydrided zircaloy 4 765wppm used to mimic high burnup effect. Cooling rate insensitive resulting phase thicknesses and their oxygen contents. Limited effect of cooling rates on residual ductility for non pre hydrided. Microstructure changes statistical variation of RCT load displacement curve.
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S0022311520310151
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Low cycle fatigue behavior of 316LN stainless steel 304 SS and 308L weld metal was investigated in air and high temperature pressurized water environment . The effects of mechanical factors and environmental factors on fatigue lives were considered . The fatigue mean curve and design curve for nuclear grade austenitic SSs in air are obtained by fitting fatigue data . The corrosion fatigue model for nuclear grade austenitic SSs in high temperature pressurized water environment is proposed based on environmental fatigue correction factor method which mainly considers the effects of strain rate temperature and dissolved oxygen concentration .
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Fatigue mean and design curves for austenitic stainless steels in air are built. High temperature water decreases fatigue life of austenitic stainless steels. Corrosion fatigue model for austenitic stainless steels is proposed.
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S0022311520310187
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Thermal conductivity and melting temperature of nuclear fuel are essential for analysing its performance under irradiation since they determine the fuel temperature profile and the melting safety margin respectively . A starting literature review of data and correlations revealed that most models implemented in state of the art fuel performance codes describe the evolution of thermal conductivity and melting temperature of Light Water Reactor MOX fuels in limited ranges of operation and without considering the complete set of fundamental dependencies . Since innovative Generation IV nuclear reactor concepts employ MOX fuel to be irradiated in Fast Reactor conditions codes need to be extended and validated for application to design and safety analyses on fast reactor MOX fuel . The aim of this work is to overcome the current modelling and code limitations providing fuel performance codes with suitable correlations to describe the evolution under irradiation of fast reactor MOX fuel thermal conductivity and melting temperature . The new correlations have been obtained by a statistically assessed fit of the most recent and reliable experimental data . The resulting laws are grounded on a physical basis and account for a wider set of effects on MOX thermal properties providing clear ranges of applicability for each parameter considered . As a first test series the new correlations have been implemented in the TRANSURANUS fuel performance code compared to state of the art correlations and assessed against integral data from the HEDL P 19 fast reactor irradiation experiment . The integral validation provides promising results pointing out a satisfactory agreement with the experimental data meaning that the new models can be efficiently applied in engineering fuel performance codes .
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Derivation of original and comprehensive correlations for thermal conductivity and melting temperature of uranium plutonium mixed oxide MOX nuclear fuel. Comparison and validation of the correlations against experimental measurements on fast reactor MOX. Simulation of fast reactor irradiation experiment with the TRANSURANUS fuel performance code. Integral validation of the new correlations against experimental power to melt data from pins in fast start up irradiation conditions. Demonstration of the efficient applicability of the correlations in engineering fuel performance codes.
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S0022311520310205
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The mechanical behavior microstructure and texture evolution were investigated during isothermal annealing at 1300 C of pure tungsten moderately warm rolled to 67 thickness reduction . The degradation of the mechanical properties is characterized by hardness testing . The microstructure and texture evolution during heat treatment were characterized by Electron Backscatter Diffraction . During annealing of the moderately warm rolled tungsten recrystallization occurred first quickly followed by relatively slow grain growth . The recrystallized volume fractions determined from hardness measurements and microstructural characterization were essentially the same . The evolution of the grain sizes during recrystallization was analyzed independently for deformed and recrystallized grains . Quantitative texture analysis showed that the overall texture strength is enhanced after recrystallization . As recrystallization strongly affects the mechanical properties of tungsten such insights in the annealing behavior of warm rolled tungsten plates are valuable for an understanding of their performance as potential plasma facing materials in future fusion reactors .
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Microstructural evolution during annealing of pure tungsten warm rolled by 67 . ESBD reveals fast discontinuous recrystallization followed by slower grain growth. Recrystallization sharpens grain size distribution scaling during grain growth. Recrystallized grains keep aspect ratio deformed grains show transient increase. Texture strength increases during recrystallization decreases during grain growth.
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S0022311520310217
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A high intensity proton beam exposure with 181 MeV energy has been conducted at Brookhaven Linac Isotope Producer facility on various material specimens for accelerator targetry applications including titanium alloys as a beam window material . The radiation damage level of the analyzed capsule was 0.25 dpa at beam center region with an irradiation temperature around 120 C. Tensile tests showed increased hardness and a large decrease in ductility for the dual phase Ti 6Al 4V Grade 5 and Grade 23 extra low interstitial alloys with the near phase Ti 3Al 2.5V Grade 9 alloy still exhibiting uniform elongation of a few after irradiation . Transmission Electron Microscope analyses on Ti 6Al 4V indicated clear evidence of a high density of defect clusters with size less than 2nm in each phase grain . The phase grains did not contain any visible defects such as loops or black dots while the diffraction patterns clearly indicated phase precipitation in an advanced formation stage . The radiation induced phase transformation in the phase could lead to greater loss of ductility in Ti 6Al 4V alloys in comparison with Ti 3Al 2.5V alloy with less phase .
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High intensity proton beam irradiation on dual phase titanium alloys up to 0.25 dpa. Modest work hardening capability remains for Ti 3Al 2.5V while none for Ti 6Al 4V. High density defect clusters in grains and nano scale phase precipitation in. Radiation induced phase could lead to greater loss of ductility for Ti 6Al 4V
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S0022311520310242
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The lifetime of tungsten monoblocks under fusion conditions is ambivalent . In this work the microstructure dependent mechanical behaviour of pulsed high heat flux exposed W monoblock is investigated . Two different microstructural states i.e . initial and recrystallized both machined from HHF exposed monoblocks are tested using tensile and small punch tests . The initial microstructural state reveals a higher fraction of low angle boundaries along with a preferred orientation of crystals . Following HHF exposure the recrystallized state exhibits weakening of initial texture along with a higher fraction of high angle boundaries . Irrespective of the testing methodology both the microstructural states display brittle failure for temperatures lower than 400
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Microstructure dependent high temperature mechanical testing of HHF exposed W monoblocks. EBSD was used to distinguish the microstructural states and for further correlation with the mechanical behaviour. Irrespective of the testing methodology temperature the recrystallized state showed better ductility in comparison to the initial deformed state. Recrystallization assisted lowering of brittle to ductile transition temperature BDTT occurred.
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S0022311520310254
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Fusion energy has great potential over other sources of energies due to the abundance of fusion fuel on the Earth and tokamak has turns out to be the best technique to harvest fusion energy . However a continuous operation of a tokamak involves challenges due to plasma wall interaction e.g . erosion re deposition fuel retention and impurities control . For an efficient operation the tokamak wall has to be remotely monitored and laser induced breakdown spectroscopy seems to be the most suitable technique for this purpose . Multiple LIBS techniques for this purpose are reviewed in this contribution . The role of pressure excitation wavelength and atmospheric effect has also been discussed . LIBS studies of spatial and depth profiles of W Be Al Mo Li C based materials with impurities are presented as well and compared to other analytical methods . At last the measurement of layer wise matrix hardness is also discussed .
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PFCs are crucial parts of the fusion machines and LIBS is a robust methodology for its study due to its unique merits. The investigation of impurities anti corrosion materials and the diagnostics of fuel retention on the PFCs is presented. Conventional back collection and dual pulse LIBS methods their merits and requirement for fusion research are explained. Identification of elements and molecular band spatial and depth profile and hardness analysis by LIBS are discussed.
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S0022311520310266
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Estimating the lifetime of W armoured divertor components is of great importance for ITER . During heat loading one of the mechanisms inducing tungsten microstructure modification is recrystallization . As tungsten recrystallization induces a decrease of the mechanical properties the knowledge of the recrystallization kinetics of the ITER tungsten material is necessary over the entire divertor PFU operational temperature window . Some data exist for temperature loading lower than 1350C . For higher temperatures some constraints due to the heating system and its impact on recrystallization process limit the possibility of high temperature studies . For this reason in this study a laser heating device and related diagnostics are used for the determination of the recrystallization kinetics . The Johnson Mehl Avrami Kolmogorov equation is used to model the recrystallization process . Recrystallization kinetics for annealing temperatures higher than 1400C are provided for two tungsten materials produced according to the ITER specifications . At 1600C recrystallization fraction of 50 is obtained after 200s thermal loading for one batch and after 500s for the other one .
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Recrystallization kinetics of temperature processed tungsten is studied for the first time above 1350C and up to 2000C. Recrystallization fractions of tungsten batches from two manufacturers were studied. At 1600C recrystallization fraction of 50 is reached after 200s thermal loading for one batch and after 500s for the other one.
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S0022311520310357
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Uranium rich UZrNb and UZrMo alloys are considered to be good alternative fuel for fast reactors . From that prospective number of selected uranium rich binary and ternary U6X alloy samples with total alloying content of 6 wt were prepared and characterized by X ray diffraction scanning electron microscope equipped with energy dispersive spectroscope and differential thermal analyzer . Molar heat capacity is considered as one of the most important thermophysical properties of nuclear fuel . The molar heat capacity measurements of U6Zr U2Zr4Nb U6Nb U4Zr2Mo and U2Zr4Mo alloys were carried out using differential scanning calorimeter in the temperature range of 323823K . The U6Zr alloy is considered here as the reference composition and the effect of Nb or Mo addition on the molar heat capacity data of U6X alloys has been highlighted . The thermodynamic functions such as molar enthalpy increment
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Measurement of molar heat capacity of U6 Zr Nb and U6 Zr Mo alloys using DSC. Molar heat capacity decreases with addition of 6wt alloying element Zr Nb Mo . Replacement of Zr in UZr alloy with Nb Mo reduces molar heat capacity. Measured molar heat capacities are lower than that estimated by Neumann Kopp rule. Evaluation of molar enthalpy and entropy increments using molar heat capacity.
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S002231152031062X
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The structural changes of ammonium diuranate microspheres prepared by the solgel route via internal gelation were investigated during thermal treatments in oxidative and reducing conditions . In particular
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Crack formation during calcination in air observed cracks heal and spheres shrink. Fractures and shrinkage differs significantly for Ce. compared to Ce. precursor. Macroscopic behaviour related to release of volatile decomposition products. Less rapid reduction of. matrix to UO. None. for Ce. precursor Ar H. 700C . Shrinkage mainly attributed to sintering effects and less to phase transitions.
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S0022311520310692
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The Fe13Cr4.5Al2Mo alloys one of the most promising candidates for the accident tolerant fuel cladding in the light water reactors were strengthened with Nb addition and the strengthening mechanism were assessed by an internal friction measurement method combined with microstructure characterization . The FeCrAlMo alloy with 1.0wt Nb addition exhibited the highest ultimate tensile strength and an acceptable ductility at each tested temperature . For example at 700C the UTS of FeCrAlMo 1wt Nb was about 215MPa which was 53.5 higher than that of Nb free FeCrAlMo alloy . For Nb free FeCrAlMo alloy a high IF peak P
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FeCrAl alloys with 1.0wt Nb addition exhibit the highest mechanical properties. Strengthening mechanism is assessed by internal friction measurement. High density nanoscale Fe. Nb Laves particles precipitate on the grain boundaries. High density nanoscale Fe. Nb Laves particles can also precipitate within the matrix. Dislocation grain boundary cannot move below 800C because of pinning of Fe. Nb phase.
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S0022311520310710
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A micromechanical model for quantifying the simultaneous influence of irradiation hardening and swelling on the mechanical stiffness and strength of neutron irradiated austenitic stainless steels is proposed . The material is regarded as an aggregate of equiaxed crystalline grains containing a random dispersion of pores and exhibiting elastic isotropy but viscoplastic anisotropy . The overall properties are obtained via a judicious combination of various bounds and estimates for the elastic energy and viscoplastic dissipation of voided crystals and polycrystals . Reference results are generated with full field numerical simulations for dense and voided polycrystals with periodic microstructures and crystal plasticity laws accounting for the evolution of dislocation and Frank loop densities . These results are calibrated with experimental data available from the literature and are employed to assess the capabilities of the proposed model to describe the evolution of mechanical properties of highly irradiated Solution Annealed 304L steels at 330
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Stiffness and strength of neutron irradiated austenitic steels are studied. Simultaneous influence of irradiation hardening and swelling is studied. The material is seen as polycrystal with voids due to large irradiation levels. A micromechanical model is proposed for an elasto viscoplastic description. v A decrease of overall elastic properties and strength with porosity is reported.
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