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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 685 records · Page 38
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Fast, quantitative, and nondestructive evaluation of hydrided LWR fuel cladding by small angle incoherent neutron scattering of hydrogen

A non-destructive neutron scattering method to precisely measure the uptake of hydrogen and the distribution of hydride precipitates in light water reactor (LWR) fuel cladding was developed. Zircaloy-4 cladding used in commercial LWRs was used to produce hydrided specimens. The hydriding apparatus consists of a closed stainless steel vessel that contains Zr alloy specimens and hydrogen gas. Following hydrogen charging, the hydrogen content of the hydrided specimens was measured using the vacuum hot extraction method, by which the samples with desired hydrogen concentration were selected for the neutron study. Optical microscopy shows that our hydriding procedure results in uniform distribution of circumferential hydrides across the wall. Small angle neutron incoherent scattering was performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. This study demonstrates that the hydrogen in commercial Zircaloy-4 cladding can be measured very accurately in minutes by this nondestructive method over a wide range of hydrogen concentrations from a very small amount ( 20 ppm) to over 1000 ppm. The hydrogen distribution in a tube sample was obtained by scaling the neutron scattering rate with a factor determined by a calibration process using standard, destructive direct chemical analysis methods on the specimens. This scale factor will be used in future tests with unknown hydrogen concentrations, thus providing a nondestructive method for absolute hydrogen concentration determination.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Crystallographic texture and microstructural changes in fusion welds of recrystallized Zry-4 rolled plates

This study presents a detailed characterization of the microstructural and crystallographic texture changes observed in the transition region in a weld between two Zircaloy-4 cold rolled and recrystallized plates. The microstructural study was performed by optical microscopy under polarized light and scanning electron microscopy (SEM). Texture changes were characterized at different lengthscales: in the micrometric size, orientation imaging maps (OIM) were constructed by electron backscatter diffraction (EBSD), in the millimetre scale, high energy XRD experiments were done at the Advanced Photon Source (USA) and compared to neutron diffraction texture determinations performed in the HIPPO instrument at Los Alamos National Laboratory. In the heat affected zone (HAZ) we observed the development of Widmanstätten microstructures, typical of the α( hcp ) to β( bcc ) phase transformation. Associated with these changes a rotation of the c-poles is found in the HAZ and fusion zone. While the base material shows the typical texture of a cold rolled plate, with their c-poles pointing 35° apart from the normal direction of the plate in the normal-transversal line, in the HAZ, c-poles align along the transversal direction of the plate and then re-orient along different directions, all of these changes occurring within a lengthscale in the order of mm. The evolution of texture in this narrow region was captured by both OIM and XRD, and is consistent with previous measurements done by Neutron Diffraction in the HIPPO diffractometer at Los Alamos National Laboratory, USA. The microstructural and texture changes along the HAZ were interpreted as arising due to the effect of differences in the cooling rate and β grain size on the progress of the different α variants during transformation. Fast cooling rates and large β grains are associated to weak variant selection during the β–>α transformation, while slow cooling rates and fine β grains result in strong variant selection.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Evaluating the irradiation effects on the elastic properties of miniature monolithic SiC tubular specimens

The initial results of a post-irradiation examination study conducted on CVD SiC tubular specimens irradiated under a high radial heat flux are presented herein. The elastic moduli were found to decrease more than that estimated based on previous studies. The significant decreases in modulus are attributed to the cracks present in the specimens. The stresses in the specimens, calculated through finite element analyses, were found to be greater than the expected strength of irradiated specimens, indicating that the irradiation-induced stresses caused these cracks. In conclusion, the optical microscopy images and predicted stress distributions indicate that the cracks initiated at the inner surface and propagated outward.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Nondestructive hydrogen analysis of steam-oxidized Zircaloy-4 by wide-angle neutron scattering

In this study, a nondestructive neutron scattering method to precisely measure the hydrogen content in high-temperature steam-oxidized Zircaloy-4 cladding was developed. Zircaloy-4 cladding was used to produce hydrided specimens with hydrogen content up to ≈500 wppm. Following hydrogen charging, the hydrogen content of the hydrided specimens was measured using the vacuum hot extraction method, by which the samples with desired hydrogen concentrations were selected for the neutron study. The hydrided samples were then oxidized in steam up to ≈6.0wt. % at 1100°C. Optical microscopy shows that our hydriding procedure results in uniform distribution of circumferential hydrides across the wall thickness, and uniform oxide layers were formed on the sample surfaces by the steam oxidation. Small- and wide-angle neutron scattering were simultaneously performed to provide a quick (less than an hour per sample) measurement of the hydrogen content in various types of hydrided and oxidized Zircaloy-4. Our study demonstrates that the hydrogen in pre-oxidized Zircaloy-4 cladding can be measured very accurately by both small- and wide-angle neutron scattering. For steam-oxidized samples, the small-angle neutron scattering is contaminated with coherent scattering from additional structural features induced by the steam oxidation. However, the scattering intensity of the wide-angle neutron scattering increases proportionally with the hydrogen charged in the samples. The hydrogen content and wide-angle neutron scattering intensity are highly linearly correlated for the oxidized cladding samples examined in this work, and can be used to precisely determine the hydrogen content in steam-oxidized Zircaloy-4 samples. Hydrogen contents determined by neutron scattering of oxidation samples were also found to be consistent with the results of chemical analysis within acceptable margins for error.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Postirradiation examinations of low burnup U 3 Si 2 fuel for light water reactor applications

The United States Department of Energy Advanced Fuels Campaign in collaboration with industry is working to develop enhanced accident tolerant fuels for light water reactors. The initial irradiations of these concepts were performed in the Idaho National Laboratory Advanced Test Reactor using drop-in style irradiations. These irradiations are collectively referred to as Accident Tolerant Fuels 1 (ATF-1). The postirradiation examination of ATF-1 has begun and some concepts have now completed initial examination. This work is focused on postirradiation examination data of U 3 Si 2 fuels at low burnup (i.e., < 20 GWd/tHM) for application in Light Water Reactors (LWRs). The U 3 Si 2 pellets show limited cracking in comparison to the expected behavior of UO 2 at same power level. Moreover, gamma spectrometry data did not reveal migration of fission products. Minor homogeneous hardening along the pellet radius due to accumulation of fission products and radiation damage was measured by microindentation. The fission gas release and swelling remain very low. Formation of fission gas bubbles resolvable with optical microscopy occurs from the pellets center outward to approximately 60% of the fuel pellet radius. With this being said, the overall data suggest a good performance of this accident tolerant fuel candidate at low burnup.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

On residual stress analysis and microstructural evolution for stainless steel type 304 spent nuclear fuel canisters weld joint: Numerical and experimental studies

Chloride-induced stress corrosion cracking (CISCC) in the weldments of austenitic stainless steel canisters is one of the primary safety concerns during the dry storage of used nuclear fuel at Independent Spent Fuel Storage Installations in coastal areas. In order to evaluate the CISCC behavior in the canister, a 3D sequentially coupled thermo-mechanical finite element model was built to simulate the residual stresses induced by two intersecting longitudinal and circumferential multi-pass welds in austenitic stainless steel mockup canister. Weld-induced residual stresses from simulation were in good agreement with the experimental measurement results by deep-hole drilling and contour methods. Through-wall high tensile axial and hoop stresses were observed in both the fusion zone and heat affected zone (HAZ) in the longitudinal and circumferential weld, respectively. On the other hand, the microstructure within the weldments of 304 stainless steel mockup was characterized using optical microscopy, scanning electron microscopy, electron back-scattered diffraction and energy dispersive X-ray spectroscopy methods. A few amount of MnS inclusions in both the base metal and HAZ, and a lot of (Mn,Si) x O y inclusions in the fusion zone were found. Meanwhile, there is some α'-martensite in the base metal, while the martensite was reversed in the HAZ. In the fusion zone, there is a lot of δ-ferrite. The martensite reversion phenomenon was also verified by the combination of Feritscope test and thermal simulation, which could predict a martensite reversion temperature as 400C°. The welding residual stresses and microstructure analysis results could offer the instructive information for the prediction of CISCC behavior of the spent nuclear fuel canisters.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Method for measurement of TRISO kernel and layer volumes by X-ray computed tomography

Layer dimensions are key parameters for as-fabricated tristructural-isotropic (TRISO) particle fuel as well as for post-irradiation examination of particle performance. Layer thicknesses are typically measured by optical microscopy of the particle cross section near mid-plane, while layer volumes are estimated with serial sectioning and microscopy. This method for measuring layer volumes is limited due to the resolution limit imposed by slice thickness and the effect of the polishing process on delicate irradiated particle microstructures. In this study, image processing software has been developed to segregate the three-dimensional (3D) TRISO particle images provided by x-ray computed tomography (XCT) into high-resolution volumetric data for the kernel, all particle layers, and any internal voids introduced during irradiation. These data can be used to analyze key in-reactor behaviors of TRISO particles such as kernel swelling or buffer shrinkage, which are important inputs for fuel performance modeling.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Microstructural evolution of Mo-UO 2 cermets under high temperature hydrogen environments

Ceramic-metallic (cermet) materials show promise for use in nuclear thermal propulsion applications due to attractive thermophysical properties including high temperature stability and high thermal conductivity. In this work, molybdenum-uranium dioxide (Mo-UO 2 ) cermet fuel elements were fabricated by means of spark plasma sintering (SPS) and were subsequently exposed to hydrogen at high temperatures (2500 K). Mo-UO 2 samples pre- and post-exposure were characterized by means of optical microscopy, scanning electron microscopy, and X-ray diffraction (XRD). Microscopy analyses of the as-produced material displayed microscopic cracking on the interior of the spherical UO 2 fuel particles but confirmed that the fuel particles were fully encapsulated in the Mo matrix. The results further showed mass loss, macroscopic swelling, and cracking in the cermet samples which occurred during high temperature hydrogen testing. Nanoscale swelling was evidenced by XRD in the Mo matrix and UO 2 fuel structure due to the incorporation of defects and accompanied microstrain.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Interaction of beryllium with 316H stainless steel in molten Li 2 BeF 4 (FLiBe)

The increased demand in renewable energy resources has led to a renewed interest in nuclear reactors including a new generation of molten salt reactors. One of the issues encountered in handling of molten chloride or fluoride salts is the inability of metals to form and retain a protective oxide surface layer, generally chromium oxide. The low chromium content, nickel-based alloy INOR-8, later marketed as Hastelloy N, was specifically developed to contain molten fluorides. This alloy had some deficiencies, but its composition resulted in it incurring far less chromium depletion by the molten fluoride salts than alloys like 316 stainless steel. For the next generation of molten salt reactors, an alloy with pressure vessel code approval and higher temperature capability than Hastelloy N is required. Consequently, several reactor designers have reportedly chosen to build the reactor containment vessel from 316H stainless steel. In order to minimize corrosion of the stainless steel by the molten fluoride salt (2LiF-BeF 2 ), it is proposed to add beryllium metal to the salt to react with impurities and provide a means to lower the oxidation potential thus making the salt less corrosive. Even so, there is a concern whether the beryllium would react with components of the stainless steel. To address this concern, static capsule tests were conducted in which selected amounts of beryllium were added to capsules containing 2LiF-BeF 2 (FLiBe) salt and 316H stainless steel and Hastelloy N samples. Following exposure, the samples were cleaned and examined using a variety of analysis techniques including, optical microscopy, SEM-EDS, XPS, LIBS, XRD and EPMA. Tensile samples were also exposed in capsules, and those were subjected to tensile testing. Notably, It was found that for the level of beryllium additions used in this study, intermetallic compounds were formed which could be detrimental to the long-term performance of the 316H stainless steel.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Radiation-induced alteration of sandstone concrete aggregate

This study investigates the alteration of felsic sandstone-type rock, which is used as a coarse aggregate in concrete, subject to the effects of gamma-ray and neutron irradiation. The effects of three gamma-ray doses (27, 55, and 108 MGy) and four neutron fluence levels (1.22, 2.19, 6.99, and 14.30 × 10 19 n/cm 2 , $E$ ≥ 0.01 MeV) were investigated. Quartz and albite were found to be the major rock-forming minerals, with microcline intermediates, chlorite, and muscovite as the minors. Gamma rays caused no significant changes to the physical properties of the sandstone aggregates, even at high doses (108 MGy). In contrast, neutron irradiation caused alterations that became more pronounced at higher neutron fluences. The solid was confirmed to expand through metamictization of the rock-forming minerals. Quartz and muscovite were the most affected phases, whereas albite and microcline intermediates were only slightly affected, and chlorite was almost unaffected. The decrease in density was measured by He and water pycnometry, and this value was almost reproduced by calculations using the rock-forming mineral composition of the pristine sample measured using X-ray powder diffraction/Rietveld analysis and the cell volume change of the major forming minerals. In addition, light optical microscopy and scanning electron microscopy images confirmed the presence of intergranular and intragranular cracks. Intergranular cracks appeared to have initiated from the quartz grains, which expanded significantly. The intragranular cracks were frequently observed in the albite and microcline intermediates. These cracks can be described as radial cracks starting from the expanding quartz, caused by enforced displacement for deformation consistency with quartz expansion. The crack area ratio quantified by SEM image analysis corresponds to the discrepancy of the volume expansion difference calculated by He or water pycnometry and dimensional change measurements. In conclusion, an evaluation of solid expansion and crack openings in aggregates is important to estimate concrete degradation.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Phase stability and microstructure of neutron-irradiated substoichiometric yttrium dihydrides

The impact of the neutron-displacement damage on phase stability and microstructure of substoichiometric yttrium dihydrides (YH x , x <2) were investigated to assess their use as solid moderator in high-temperature nuclear reactors. YH x specimens were, thus, subjected to neutron irradiations in the range of 0.1–2 displacements per yttrium atom (dpa-Y) in the temperature range of 536–878°C at the Oak Ridge National Laboratory's (ORNL's) High Flux Isotope Reactor (HFIR). YH x specimens were initially prepared at stoichiometry (H/Y) ratios of 1.69 and 1.83. HFIR-irradiated specimens were characterized by variety of techniques to investigate H retention characteristics including dimensional analysis, optical microscopy, scanning electron microscopy electron back scatter diffraction (EBSD), transmission electron microscopy, thermal desorption spectroscopy (TDS), and high-energy x-ray diffraction (HE-XRD) characterizations. Overall, YH x exhibited notable structural and phase stability under short-term neutron-irradiation, except for the samples with significant silicon carbide (SiC) interaction at high doses and temperatures. Basic dimensional and mass measurements were misleading for accurate assessment of H retention, as confirmed by EBSD phase maps, XRD line profiles, and TDS signals. Thus, it was discussed that a robust H retention metric is needed to assess irradiated hydrides. Further, nanoscale cavities were observed as a result of the neutron irradiation in all samples. Although no clear impact of dose and irradiation temperature was determined, the initial H/Y ratio had an impact on the cavity number density where low H/Y specimens had high-resistance to cavity formation. The Y-vacancy cluster formation at the collision stage of the displacement cascade and their stabilization by H were considered to be the likely underlying mechanisms for the observed cavity microstructure.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Evaluation of corrosion coupons exposed to molten-chloride electrorefiner salts long term

Corrosion mitigation has long been considered a challenge in long-term service of both pyrochemical reprocessing and molten salt reactor applications. Corrosion coupons consisting of 2.25Cr 1Mo ASME SA387 Grade 22 Class 2 steel, were installed in a uranium electrorefiner containing molten chloride salt at 500°C. This electrorefiner went into service at Idaho National Laboratory in 1994. Seven of these coupons were removed in 2019 (i.e., submerged in molten salt for 25 years) for analysis and characterization including neutron imaging, optical microscopy, scanning electron microscopy, and wave/energy dispersive x-ray spectroscopy. The results showed that all seven coupons experienced between 30 to 175 µm of surface degradation without evidence of brittle fracture. Meaning there was no evidence of accelerated corrosion or fractures, and the average steel surface loss was of less than 5 µm per year. All indications were that the environmental conditions inside the electrorefiner were favorable for the steel alloy.

36 - MATERIALS SCIENCE↗
DOE OSTI2017

Enabling aqueous processing for crack-free thick electrodes

Aqueous processing of thick electrodes for Li-ion cells promises to increase energy density due to increased volume fraction of active materials, and to reduce cost due to the elimination of the toxic solvents. Here in this paper this work reports the processing and characterization of aqueous processed electrodes with high areal loading and associated full pouch cell performance. Cracking of the electrode coatings becomes a critical issue for aqueous processing of the positive electrode as areal loading increases above 20–25 mg/cm 2 (~4 mAh/cm 2 ). Crack initiation and propagation, which was observed during drying via optical microscopy, is related to the build-up of capillary pressure during the drying process. The surface tension of water was reduced by the addition of isopropyl alcohol (IPA), which led to improved wettability and decreased capillary pressure during drying. The critical thickness (areal loading) without cracking increased gradually with increasing IPA content. The electrochemical performance was evaluated in pouch cells. Electrodes processed with water/IPA (80/20 wt%) mixture exhibited good structural integrity with good rate performance and cycling performance.

25 ENERGY STORAGE↗
DOE OSTI2020

Impact of electrode thick spot irregularities on polymer electrolyte membrane fuel cell initial performance

Polymer electrolyte membrane fuel cells have great potential for power generation in many applications, but to implement them on a larger scale, understanding of quality requirements, tolerances, and controls during manufacturing must improve. As part of this broad challenge, irregularities resulting from the membrane electrode assembly production process must be studied to determine if they impact the performance of the cell. One potential irregularity is a locally increased thickness of the coated or deposited electrode. In order to understand and provide insight into the impact of such irregularities, we intentionally created thick spots in the center of cathode electrodes. The thick spots fabricated by two different approaches were carefully characterized by optical microscopy, X-ray fluorescence spectroscopy and scanning electron microscopy (SEM). We found that the impact of the thick spots on the electrochemically active catalyst area and mass activity was minimal. However, specific electrode thick spot irregularities significantly impacted the mass transport properties of the cells. Based on SEM analysis, the likely cause of the performance drop is morphological changes associated with the creation of the thick spot irregularity. Spatial performance results obtained with a segmented cell system further indicated that the area of the performance impact exceeded that of the intentionally created thick spot.

25 ENERGY STORAGE↗
DOE OSTI2023

The effect of ink ball milling time on interparticle interactions and ink microstructure and their influence on crack formation in rod-coated catalyst layers

This work investigates the influence of ballmilling (sometimes also referred to as jar roller milling) time on cathode catalyst layer (CL) inks and electrode properties using formulations and coating methods relevant for industrial manufacturing. Four CL inks with the same composition were milled for 24, 48, 72, or 96 h. Rheological investigation of these inks showed a reduction of elastic moduli and steady-shear viscosity with continuous ink milling, which is correlated to a decrease in particle-particle interactions as well as formation of smaller agglomerates. Optical microscopy (OM) analysis of the fabricated electrodes revealed a trend in surface crack formation; formulations milled for 24 h contained the lowest average surface crack area percentages of 0.370% at heavy-duty loadings of ~0.300 mg Pt cm -2 , compared to 2.418% for the ink milled for 96 h. Further characterization of the CL through transmission electron microscopy (TEM) imaging showed a decrease in the mean agglomerate and pore size with milling time. Furthermore, these smaller electrode features were consistent with reduced fracture resistance and, hence, development of larger stresses during drying. Our results highlight the need to consider ink processing as an important component in defect-free CL manufacturing.

30 DIRECT ENERGY CONVERSION↗
DOE OSTI2014

Faulting processes in active faults - Evidences from TCDP and SAFOD drill core samples

The microstructures, mineralogy and chemistry of representative samples collected from the cores of the San Andreas Fault drill hole (SAFOD) and the Taiwan Chelungpu-Fault Drilling project (TCDP) have been studied using optical microscopy, TEM, SEM, XRD and XRF analyses. SAFOD samples provide a transect across undeformed host rock, the fault damage zone and currently active deforming zones of the San Andreas Fault. TCDP samples are retrieved from the principal slip zone (PSZ) and from the surrounding damage zone of the Chelungpu Fault. Substantial differences exist in the clay mineralogy of SAFOD and TCDP fault gouge samples. Amorphous material has been observed in SAFOD as well as TCDP samples. In line with previous publications, we propose that melt, observed in TCDP black gouge samples, was produced by seismic slip (melt origin) whereas amorphous material in SAFOD samples was formed by comminution of grains (crush origin) rather than by melting. Dauphiné twins in quartz grains of SAFOD and TCDP samples may indicate high seismic stress. The differences in the crystallographic preferred orientation of calcite between SAFOD and TCDP samples are significant. Microstructures resulting from dissolution–precipitation processes were observed in both faults but are more frequently found in SAFOD samples than in TCDP fault rocks. As already described for many other fault zones clay-gouge fabrics are quite weak in SAFOD and TCDP samples. Clay-clast aggregates (CCAs), proposed to indicate frictional heating and thermal pressurization, occur in material taken from the PSZ of the Chelungpu Fault, as well as within and outside of the SAFOD deforming zones, indicating that these microstructures were formed over a wide range of slip rates.

Janssen, C.↗
DOE OSTI2011

Effect of the surface preparation techniques on the EBSD analysis of a friction stir welded AA1100-B{sub 4}C metal matrix composite

Aluminum based metal matrix composites (MMCs) have been used in various automobile, aerospace and military industries. Yet characterization of the microstructure in these materials remains a challenge. In the present work, the grain structure in the matrix of B{sub 4}C particulate reinforced MMCs and their friction stir welds is characterized by using optical metallography and the electron backscatter diffraction (EBSD) technique. Optical metallography can partially reveal the grain structure in the matrix of AA1100-16 vol.% B{sub 4}C composite. The EBSD technique has been successfully applied to characterize the grain structure in the AA1100-16 vol.% B{sub 4}C friction stir welds, which provides a powerful tool to follow the microstructural evolution of MMC materials during friction stir welding (FSW). Both mechanical polishing and ion beam polishing are used for the EBSD sample preparation. The effect of the sample preparation on the EBSD data acquisition quality is studied. Some typical examples, such as the identification of grains and subgrains, grain size distribution, deformation fields and the texture components are given. - Highlights: {yields} EBSD has been used to characterize the grain structure of Al-B{sub 4}C MMCs. {yields} Mechanical and ion beam polishing are compared for EBSD sample preparation of MMCs. {yields} EBSD shows great advantages over optical microscopy for microtexture analysis of MMCs.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Archaeometallurgical investigation of the iron anchor from the Tantura F shipwreck

The Tantura F shipwreck was a coaster or a fishing vessel about 15.7 m long, discovered in the Dor/Tantura lagoon, Israel in 1995. It was dated to between the mid-7th and the end of the 8th centuries CE. Among the finds excavated were two T-shaped type iron anchors. Of the two anchors, one (anchor A) was thoroughly studied by archaeometallurgical methods in order to identify forge-welding lines, to determine the welding quality and to understand the manufacturing technology. The examinations included X-ray radiography, XRF analysis, optical microscopy, SEM/EDS observation and analysis, OES analysis and microhardness tests. The investigation included characterization of the composition, microstructure, thermal treatments, forge-welding junctions and slag analysis. The results revealed a heterogeneous microstructure, rich in glassy, fayalite and wüstite slag. Iron based phases included ferrite, pearlite, cementite and Widmanstätten plates, all typical to wrought iron. The forge-welds of Anchor A were located. Each arm was made of one piece, weighing about 2.5–3 kg and the shank was made of a few 1.5–2 kg pieces. The second anchor (anchor B) was only briefly examined visually and with a few radiographs, which support the results from anchor A. The research results revealed significant information about T-shaped anchors and their manufacturing process, including hot-working processes without any additional heat treatments, and folding techniques. The microstructure was similar to other ancient simple tools such as saws, sickles, axes and mortise chisels, and though the technology to make complicated structures and objects, such as swords, existed at that time, the anchors did not require this sophistication; thus simpler techniques were used, presumably because they were more cost-effective. - Highlights: ► Tantura F was a coaster dated to mid-7th–end-8th centuries. ► Two iron anchors were discovered at the Tantura F shipwreck-site. ► Anchor A was manufactured from heterogeneous wrought iron blooms. ► Forge-welding lines were detected using archaeometallurgical methods.

36 MATERIALS SCIENCE↗