Irradiation Behavior Research Articles

Bi-model short-term solar irradiance prediction using support vector regressors

This paper proposes an accurate short-term solar irradiance prediction scheme via support vector regression. Utilizing clearness index conversion and appropriate features, the support vector regression models are able to output satisfying prediction results. The prediction results are further improved by the proposed ramp-down event forecasting and solar irradiance refinement procedures. With the help of all-sky image analysis, two separated regression models are constructed based on the cloud obstruction conditions near the solar disk. With bi-model prediction, the behavior of the changing irradiance can be captured more accurately. Moreover, if a ramp-down event is forecasted, the predicted irradiance is corrected based on the cloud cover ratio in the area near the sun. The experiments have shown that the proposed method can effectively improve the prediction accuracy on a highly challenging dataset.

Evaluating the thermal damage resistance of graphene/carbon nanotube hybrid composite coatings.

We study laser irradiation behavior of multiwalled carbon nanotubes (MWCNT) and chemically modified graphene (rGO)-composite spray coatings for use as a thermal absorber material for high-power laser calorimeters. Spray coatings on aluminum test coupon were exposed to increasing laser irradiance for extended exposure times to quantify their damage threshold and optical absorbance. The coatings, prepared at varying mass % of MWCNTs in rGO, demonstrated significantly higher damage threshold values at 2.5 kW laser power at 10.6 μm wavelength than carbon paint or MWCNTs alone. Electron microscopy and Raman spectroscopy of irradiated specimens show that the coating prepared at 50% CNT loading endure at least 2 kW.cm−2 for 10 seconds without significant damage. The improved damage resistance is attributed to the unique structure of the composite in which the MWCNTs act as an efficient absorber of laser light while the much larger rGO sheets surrounding them, dissipate the heat over a wider area.

Irradiation behaviour of α2 and γ phases in He ion implanted titanium aluminide alloy

A Ti–45Al–2Nb–2Mn + 0.8 vol.% TiB2 (at.%) alloy with fully lamellar microstructure consisting of hexagonal-close-packed (hcp) α2 and face-centred-tetragonal (fct) γ phases was irradiated by implanting helium ions to different fluences. Microstructural examination showed that helium cavities are formed in both the α2 and γ phases after He-ion irradiation. However, the helium cavities and their size change with fluence are much larger in the α2 phase than those in the γ phase, indicating that the γ phase exhibits better tolerance to the He-ion irradiation than the α2 phase. Since α2 and γ phases have different crystal structures, they possess differences in helium solubility and interstitial migration. These differences are responsible for the variation in radiation damage behaviour between the two phases.

Microstructural Evolution of an Ion Irradiated Ni–Mo–Cr–Fe Alloy at Elevated Temperatures

The irradiation behavior of a Ni­Mo­Cr­Fe alloy, of the type currently being considered for use in future molten salt cooled reactors, has been investigated in situ using 1MeV Kr ions at temperatures of 723 and 973K. When irradiated to 5 dpa, experimental observations reveal the instantaneous formation and annihilation of point defect clusters, with such processes attributed to the long range elastic interactions that occur between defects through multiple intra-cascade overlap. Corresponding differences in the defect cluster density and size distribution suggest that changes to the microstructure were dependent upon temperature and dose, affecting the growth, accumulation and mobility of irradiationinduced defect clusters under these conditions. [doi:10.2320/matertrans.MD201311]

POST-IRRADIATION ANALYSES OF U-MO DISPERSION FUEL RODS OF KOMO TESTS AT HANARO

Since 2001, a series of five irradiation test campaigns for atomized U-Mo dispersion fuel rods, KOMO-1, -2, -3, -4, and -5, has been conducted at HANARO (Korea) in order to develop high performance low enriched uranium dispersion fuel for research reactors. The KOMO irradiation tests provided valuable information on the irradiation behavior of U-Mo fuel that results from the distinct fuel design and irradiation conditions of the rod fuel for HANARO. Full size U-Mo dispersion fuel rods of 4-5 g-U/cm3 were irradiated at a maximum linear power of approximately 105 kW/m up to 85% of the initial U-235 depletion burnup without breakaway swelling or fuel cladding failure. Electron probe microanalyses of the irradiated samples showed localized distribution of the silicon that was added in the matrix during fuel fabrication and confirmed its beneficial effect on interaction layer growth during irradiation. The modifications of U-Mo fuel particles by the addition of a ternary alloying element (Ti or Zr), additional protective coatings (silicide or nitride), and the use of larger fuel particles resulted in significantly reduced interaction layers between fuel particles and Al.

An approach for evaluating the integrity of fuel applied in Innovative Nuclear Energy Systems

One of the important issues in the study of Innovative Nuclear Energy Systems is evaluating the integrity of fuel applied in Innovative Nuclear Energy Systems. An approach for evaluating the integrity of the fuel is discussed here based on the procedure currently used in the integrity evaluation of fast reactor fuel. The fuel failure modes determining fuel life time were reviewed and fuel integrity was analyzed and compared with the failure criteria.Metal and nitride fuels with austenitic and ferritic stainless steel (SS) cladding tubes were examined in this study. For the purpose of representative irradiation behavior analyses of the fuel for Innovative Nuclear Energy Systems, the correlations of the cladding characteristics were modeled based on well-known characteristics of austenitic modified 316 SS (PNC316), ferritic–martensitic steel (PNC–FMS) and oxide dispersion strengthened steel (PNC–ODS).The analysis showed that the fuel lifetime is limited by channel fracture which is a nonductile type (brittle) failure associated with a high level of irradiation-induced swelling in the case of austenitic steel cladding. In case of ferritic steel, on the other hand, the fuel lifetime is controlled by cladding creep rupture. The lifetime evaluated here is limited to 200GWd/t, which is lower than the target burnup value of 500GWd/t. One of the possible measures to extend the lifetime may be reducing the fuel smeared density and ventilating fission gas in the plenum for metal fuel and by reducing the maximum cladding temperature from 650 to 600°C for both metal and nitride fuel.

MODELING OF INTERACTION LAYER GROWTH BETWEEN U-Mo PARTICLES AND AN Al MATRIX

Interaction layer growth between U-Mo alloy fuel particles and Al in a dispersion fuel is a concern due to the volume expansion and other unfavorable irradiation behavior of the interaction product. To reduce interaction layer (IL) growth, a small amount of Si is added to the Al. As a result, IL growth is affected by the Si content in the Al matrix. In order to predict IL growth during fabrication and irradiation, empirical models were developed. For IL growth prediction during fabrication and any follow-on heating process before irradiation, out-of-pile heating test data were used to develop kinetic correlations. Two out-of-pile correlations, one for the pure Al matrix and the other for the Al matrix with Si addition, respectively, were developed, which are Arrhenius equations that include temperature and time. For IL growth predictions during irradiation, the out-of-pile correlations were modified to include a fission-rate term to consider fission enhanced diffusion, and multiplication factors to incorporate the Si addition effect and the effect of the Mo content. The in-pile correlation is applicable for a pure Al matrix and an Al matrix with the Si content up to 8 wt%, for fuel temperatures up to 200 ℃, and for Mo content in the range of 6 ? 10wt%. In order to cover these ranges, in-pile data were included in modeling from various tests, such as the US RERTR-4, -5, -6, -7 and -9 tests and Korea’s KOMO-4 test, that were designed to systematically examine the effects of the fission rate, temperature, Si content in Al matrix, and Mo content in U-Mo particles. A model converting the IL thickness to the IL volume fraction in the meat was also developed.

EBSD and TEM characterization of high burn-up mixed oxide fuel

Understanding and studying the irradiation behavior of high burn-up oxide fuel is critical to licensing of future fast breeder reactors. Advancements in experimental techniques and equipment are allowing for new insights into previously irradiated samples. In this work dual column focused ion beam (FIB)/scanning electron microscope (SEM) was utilized to prepared transmission electron microscope samples from mixed oxide fuel with a burn-up of 6.7% FIMA. Utilizing the FIB/SEM for preparation resulted in samples with a dose rate of <0.5mRem/h compared to ∼1.1R/h for a traditionally prepared TEM sample. The TEM analysis showed that the sample taken from the cooler rim region of the fuel pellet had ∼2.5× higher dislocation density than that of the sample taken from the mid-radius due to the lower irradiation temperature of the rim. The dual column FIB/SEM was additionally used to prepared and serially slice ∼25μm cubes. High quality electron back scatter diffraction (EBSD) were collected from the face at each step, showing, for the first time, the ability to obtain EBSD data from high activity irradiated fuel.

Ion irradiation of Fe-Fe oxide core-shell nanocluster films: Effect of interface on stability of magnetic properties

A cluster deposition method was used to produce films of loosely aggregated nanoclusters (NCs) of Fe core-Fe3O4 shell or fully oxidized Fe3O4. Films of these NC on Si(100) or MgO(100)/Fe3O4(100) were irradiated to 1016 Si2+/cm2 near room temperature using an ion accelerator. Ion irradiation creates structural change in the NC film with corresponding chemical and magnetic changes which depend on the initial oxidation state of the cluster. Films were characterized using magnetometry (hysteresis, first order reversal curves), microscopy (transmission electron, helium ion), and x-ray diffraction. In all cases, the particle sizes increased due to ion irradiation, and when a core of Fe is present, irradiation reduces the oxide shells to lower valent Fe species. These results show that ion irradiated behavior of the NC films depends strongly on the initial nanostructure and chemistry, but in general saturation magnetization decreases slightly.

Effects of carbide precipitate on the mechanical properties and irradiation behavior of the low activation martensitic steel

Abstract Precipitates may be induced during the processing, heat treatment and reactor operating and will play a key role in determining the properties of materials. The effects of precipitates on the mechanical properties and irradiation behavior of the low activation martensitic steel have been investigated using transmission electron microscopy combined with tensile test. Two types of China low activation martensitic (CLAM) specimens (with and without Si) were chosen as the model materials. The significant microstructural difference between the two CLAM specimens was the distribution of fine (Ta, W, V) C precipitates which can strengthen materials. Nano-indentation test revealed that the fine (Ta, W, V) C precipitates promoted by adding silicon can also decrease the irradiation hardening ratio of CLAM steel because the MC precipitates with high interface-to-volume ratio could decrease the density of the dislocation loop induced during irradiation. The interaction between precipitates and irradiation induced dislocation loops was discussed.

Advanced oxidation-resistant iron-based alloys for LWR fuel cladding

Application of advanced oxidation-resistant iron alloys as light water reactor fuel cladding is proposed. The motivations are based on specific limitations associated with zirconium alloys, currently used as fuel cladding, under design-basis and beyond-design-basis accident scenarios. Using a simplified methodology, gains in safety margins under severe accidents upon transition to advanced oxidation-resistant iron alloys as fuel cladding are showcased. Oxidation behavior, mechanical properties, and irradiation effects of advanced iron alloys are briefly reviewed and compared to zirconium alloys as well as historic austenitic stainless steel cladding materials. Neutronic characteristics of iron-alloy-clad fuel bundles are determined and fed into a simple economic model to estimate the impact on nuclear electricity production cost. Prior experience with steel cladding is combined with the current understanding of the mechanical properties and irradiation behavior of advanced iron alloys to identify a combination of cladding thickness reduction and fuel enrichment increase (∼0.5%) as an efficient route to offset any penalties in cycle length, due to higher neutron absorption in the iron alloy cladding, with modest impact on the economics.

Microstructural evolution of U(Mo)–Al(Si) dispersion fuel under irradiation – Destructive analyses of the LEONIDAS E-FUTURE plates

Several irradiation experiments have confirmed the positive effect of adding Si to the matrix of an U(Mo) dispersion fuel plate on its in-pile irradiation behavior. E-FUTURE, the first experiment of the LEONIDAS program, was performed to select an optimum Si concentration and fuel plate heat treatment parameters for further qualification. It consisted of the irradiation of 4 distinct (regarding Si content and heat treatments), full size flat fuel plates in the BR2 reactor under bounding conditions (470W/cm2 peak BOL power, ∼70% peak burn-up). After the irradiation, the E-FUTURE plates were examined non-destructively and found to have pillowed in the highest burn-up positions. The destructive post-irradiation examination confirmed that the fuel evolves in a stable way up to a burn-up of 60%235U. Even in the deformed area (pillow) the U(Mo) fuel itself shows stable behavior and remaining matrix material was present. From the calculation of the volume fractions, the positive effect of a higher Si amount added to the matrix and the higher annealing temperature can be derived.

Phase development in a U–7 wt.% Mo vs. Al–7 wt.% Ge diffusion couple

Fuel development for the Reduced Enrichment for Research and Test Reactors (RERTR) program has demonstrated that U–Mo alloys in contact with Al develop interaction regions with phases that have poor irradiation behavior. The addition of Si to the Al has been considered with positive results. In this study, compositional modification is considered by replacing Si with Ge to determine the effect on the phase development in the system. The microstructural and phase development of a diffusion couple of U–7wt.% Mo in contact with Al–7wt.% Ge was examined by transmission electron microscopy, scanning electron microscopy and energy dispersive spectroscopy. The interdiffusion zone developed a microstructure that included the cubic-UGe3 phase and amorphous phases. The UGe3 phase was observed with and without Mo and Al solid solution developing a (U,Mo)(Al,Ge)3 phase.

Investigation of the cause of peculiar irradiation behavior of 9Cr-ODS steel in BOR-60 irradiation tests

Four experimental fuel assemblies (EFAs) containing 9Cr-ODS steel cladding fuel pins were previously irradiated in the BOR-60 to demonstrate the in-reactor performance of 9Cr-ODS steel for use as fuel cladding tubes. One of the EFAs achieved the best data, a peak burn-up of 11.9at% and a neutron dose of 51 dpa, without any microstructure instability or any fuel pin rupture. On the other hand, in another EFA (peak burn-up, 10.5at%; peak neutron dose, 44 dpa), peculiar irradiation behaviors, such as microstructure instability and fuel pin rupture, occurred. Investigations of the cause of these peculiar irradiation behaviors were carried out. The detection sensitivity in an ultrasonic inspection test was shown to be low for the metallic Cr and metallic Fe inclusions. The peculiar microstructure change reappeared with high-temperature thermal-aging of the 9Cr-ODS steel containing metallic Cr inclusions. The strength and ductility of the defective part containing metallic Cr inclusions were appreciably lower than those of a standard part without the inclusions. The combined effects of matrix Cr heterogeneity (presence of metallic Cr inclusions) and high-temperature irradiation were concluded to be the main cause of the peculiar microstructure change in 9Cr-ODS steel cladding tubes in the BOR-60 irradiation tests. They contributed to the fuel pin rupture.

Irradiation behavior evaluation of oxide dispersion strengthened ferritic steel cladding tubes irradiated in JOYO

Irradiation behavior of ODS steel cladding tubes was evaluated for the further progress in understanding of the neutron-irradiation effects on ODS steel.Two types of ODS (9Cr–ODS_F/M, 12Cr–ODS_F) steel cladding tubes with differences in basic compositions and matrix phases were irradiated in JOYO. Post-irradiation examination data concerning hardness, ring tensile property, and microstructure were obtained.Hardness measurement after irradiation showed that there was an apparent irradiation temperature dependence on hardness for 9Cr–ODS_F/M steel whereas no distinct temperature dependence for 12Cr–ODS_F steel. Also, there was no significant change in tensile strengths after irradiation below 923K, but those above 1023K up to 6.6×1026n/m2 (E>0.1MeV) were decreased by about 20%. TEM observations showed that the radiation-induced defect cluster formation during irradiation was suppressed because of high density sink site for defect such as initially-existed dislocation, and precipitate interfaces. In addition, oxide particles were stable up to the maximum doses of this irradiation test.

ODS cladding fuel pins irradiation tests using the BOR-60 reactor

In order to confirm the irradiation behavior of ODS steels and thus judge their applicability to fuel claddings, fuel pin irradiation tests using 9Cr and 12Cr-ODS claddings developed by JAEA were conducted to burn-up of 11.9 at% and neutron dose of 51 dpa in the BOR-60. Superior properties of the ODS claddings concerning FCCI, dimensional stability under irradiation and so on were confirmed and indicated good application prospects for high burn-up fuel. On the other hand, anomalous irradiation behaviors, fuel pin failure and the microstructure change containing coarse and irregular precipitates, occurred in a part of the fuel pin with 9Cr-ODS cladding. This paper describes evaluation of the obtained irradiation data and the investigation results into the cause of the anomalous irradiation behaviors.

Very High Laser-Damage Threshold of Polymer-derived Si(B)CN- Carbon Nanotube Composite Coatings

We study the laser irradiance behavior and resulting structural evolution of polymer-derived silicon-boron-carbonitride (Si(B)CN) functionalized multiwall carbon nanotube (MWCNT) composite spray coatings on copper substrate. We report a damage threshold value of 15 kWcm(-2) and an optical absorbance of 0.97 after irradiation. This is an order of magnitude improvement over MWCNT (1.4 kWcm(-2), 0.76), SWCNT (0.8 kWcm(-2), 0.65) and carbon paint (0.1 kWcm(-2), 0.87) coatings previously tested at 10.6 μm (2.5 kW CO2 laser) exposure. Electron microscopy, Raman spectroscopy, and X-ray photoelectron spectroscopy suggests partial oxidation of Si(B)CN forming a stable protective SiO2 phase upon irradiation.

Restructuring and redistribution of actinides in Am-MOX fuel during the first 24h of irradiation

Abstract In order to confirm the effect of minor actinide additions on the irradiation behavior of MOX fuel pellets, 3 wt.% and 5 wt.% americium-containing MOX (Am-MOX) fuels were irradiated for 10 min at 43 kW/m and for 24 h at 45 kW/m in the experimental fast reactor Joyo. Two nominal values of the fuel pellet oxygen-to-metal ratio (O/M), 1.95 and 1.98, were used as a test parameter. Emphasis was placed on the behavior of restructuring and redistribution of actinides which directly affect the fuel performance and the fuel design for fast reactors. Microstructural evolutions in the fuels were observed by optical microscopy and the redistribution of constituent elements was determined by EPMA using false color X-ray mapping and quantitative point analyses. The ceramography results showed that structural changes occurred quickly in the initial stage of irradiation. Restructuring of the fuel from middle to upper axial positions developed and was almost completed after the 24-h irradiation. No sign of fuel melting was found in any of the specimens. The EPMA results revealed that Am as well as Pu migrated radially up the temperature gradient to the center of the fuel pellet. The increase in Am concentration on approaching the edge of the central void and its maximum value were higher than those of Pu after the 10-min irradiation and the difference was more pronounced after the 24-h irradiation. The increment of the Am and Pu concentrations due to redistribution increased with increasing central void size. In all of the specimens examined, the extent of redistribution of Am and Pu was higher in the fuel of O/M ratio of 1.98 than in that of 1.95.

Carbothermic Reduction of Vanadium Titanomagnetite by Microwave Irradiation and Smelting Behavior

The recovery rate of the titanium is much low in the current process of the vanadium titanomagnetite resource in Panzhihua, China. The carbothermic reduction process in the microwave filed was proposed because the problem caused by Ti(C,N) should be avoid and the TiO 2 content in the slag should increase. The reduction behavior of the ore in microwave and the distribution of the elements during the smelting process were studied. The heating rate of the vanadium titanomagnetite in the microwave field can be divided into three stages: the very fast heating period in initial 25 minutes in which no apparent reduction happens, the slow heating period in which the solid reduction of the magnetite decreases the heating rate, another fast heating period in which the reduction of ilmenite happens. About 60% of iron from 1 kg of the ore and carbonaceous materials was obtained by an energy supply of 10 kW by microwave irradiation in 120 min. The recovery ratios of Fe, V and Si increase with increasing the reduction temperature. The addition of the CaO can increase the recovery ratio of Fe and V, but decrease the recovery ratio of Si.

Assembly and Irradiation Modeling of Residual Stresses in Low-Enriched Uranium Foil-Based Annular Targets for Molybdenum-99 Production

This paper considers a composite cylindrical structure, with low-enriched uranium (LEU) foil enclosed between two aluminum 6061-T6 cylinders. A recess is cut all around the outer circumference of the inner tube to accommodate the LEU foil of open-cross section. To obtain perfect contact at the interfaces of the foil and the tubes, an internal pressure is applied to the inner tube, thereby plastically and elastically deforming it. The residual stresses resulting from the assembly process are used along with a thermal stress model to predict the stress margins in the cladding during irradiation. The whole process was simulated as a steady-state two-dimensional problem using the commercial finite element code Abaqus FEA. The irradiation behavior of the annular target has been presented, and the effect of the assembly residual stresses has been discussed.