Zirconium metal-organic framework functionalized plasmonic sensor.
Abstract not provided.
Briscoe, Jayson↗
SEARCH · Engineering Papers
Results for “ZIRCONIUM”
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Structure of zirconium-93 and zirconium-91 as shown by the reactions Zr-92(d,p)Zr-93 and Zr-92(d,t)Zr-91
Deuterons of 13-MeV incident energy were scattered from Zr-92(d,p)Zr-93. The Zr-92(d,p)Zr-93 data analysis resulted in the location of 47 levels up to an excitation energy of 4.84 MeV, and the spins of 43 of these levels were identified. Essentially all the strength of the 2d5/2, 3s1/2, 2d3/2, and 1g7/2 shells was observed; and the excitation energy of their centroids was computed to be 0.00, 1.21, 2.23, and 2.37 MeV, respectively. Also, 43 percent of the 1h11/2 strength, 21 percent of the 2f7/2 strength, and 3 percent of the 3p3/2 strength were observed. In addition, the Zr-92(d,t)Zr-91 data analysis resulted in the location of 26 levels up to an excitation energy of 4.01 MeV, and the spins of 21 of these levels were identified. Most of the expected strength of the 2d5/2 and 1g9/2 shells was obtained, and the excitation energy of their centroids was computed to be 0.31 and 3.19 MeV, respectively. In addition, six l=1 states are populated belonging to either the 2p1/2 or 2p3/2 shells.
Baron, N.↗
Synthesis and Processing of Meta-stable Negative Thermal expansion Materials Zirconium Tungstate and zirconium phosphotungstate (ZrW2O8 and ZrWP2O12).
Abstract not provided.
Gordon, Margaret Ellen↗
Removal of Tritium from the Gaseous Effluents Produced by the Recovery of Zirconium from Used Fuel Cladding
U.S. regulations will require the removal of tritium ({sup 3}H) from gaseous effluents associated with nuclear fuel reprocessing and recycling. One process that may generate {sup 3}H-containing gaseous effluents is the recovery of zirconium from used fuel cladding. A chlorination process developed by Collins et al. is designed to convert the zirconium present in Zircaloy{sup R} cladding to recoverable ZrCl{sub 4}. The greatest economic benefit from including zirconium recycle as part of a larger used nuclear fuel (UNF) recycle effort would be a decrease in the disposal cost of used fuel cladding. Used fuel cladding is considered high-level waste (HLW). As zirconium is the second largest element by mass (25%) contained in most UNF assemblies (with uranium being the largest), recovery of the zirconium could result in a substantial decrease in the amount of HLW intended for geologic disposition. The reclaimed ZrCl{sub 4} product, if sufficiently free from radioactive impurities, could then be disposed of as low-level waste or reused in nuclear applications. The chlorination process uses chlorine gas (Cl{sub 2}) and an inert carrier such as argon (Ar) to convert the zirconium present in cladding to ZrCl{sub 4}. ZrCl{sub 4} is more volatile than other components in the cladding, causing evaporation of a relatively pure gaseous ZrCl{sub 4} stream. The product ZrCl{sub 4} is recovered as a solid salt by condensation when cooled to less than about 300 deg. C. A fraction of the {sup 3}H (0-96%) produced in nuclear fuel during irradiation is found in zirconium-based cladding, possibly as zirconium hydride (the wide range is a remnant of various researchers' analyses of cladding of various irradiations and cooling times). Heating the cladding to high temperatures, such as would be done during the chlorination process, will release the {sup 3}H. At the temperatures of the zirconium recycle process, the {sup 3}H is likely to react with excess Cl{sub 2} to produce tritiated HCl ({sup 3}HCl), although this could be avoided if the cladding and fuel were treated to remove {sup 3}H prior to chlorination. A previous analysis showed that the concentration of {sup 3}HCl in the off-gas of the zirconium recycle process will be in the range of parts per billion to a few parts per million. Should {sup 3}H mitigation be required for the zirconium recycle process, it will be completed through the abatement of {sup 3}HCl from this off-gas stream. The intent of the testing described in this summary was to conduct proof-of-principle demonstrations for two {sup 3}HCl removal methods. The zirconium recycle off-gas stream differs from many industrial HCl-containing streams in that the concentration of HCl requiring removal is extremely low. Demonstrations of HCl removal at ppm concentrations were not found in the literature, making the proof-of-principle approach attractive as a way to identify the most promising method for continued development. An important consideration throughout these experiments was that current research efforts on the zirconium recycle process are conducted in a hot cell and any {sup 3}HCl removal methods should be easily deployed in a hot-cell environment.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Layered zirconium phosphonate with inorganic–organic hybrid structure: Preparation and its assembly with DNA
An aminoethoxy-functionalized zirconium phosphonate (Zr(O{sub 3}POCH{sub 2}CH{sub 2}NH{sub 2}){sub 2}·3H{sub 2}O), abbreviated as ZrRP (R=OCH{sub 2}CH{sub 2}NH{sub 2}), with layered structure has been synthesized. This layered compound possesses the characteristic of inorganic–organic hybrid, due to the covalently linked aminoethoxy in the host layer. The anion exchanged property of this zirconium phosphonate is suitable for the direct intercalation of negatively charged DNA, which is different from these reported zirconium phosphates or zirconium phosphonates. As a precursor, this prepared zirconium phosphonate was utilized to fabricate a novel DNA/ZrRP binary hybrid via a delamination-reassembly procedure. The release behavior of DNA from the DNA/ZrRP composite was investigated at different medium pH, because the combination between zirconium phosphonate sheets and DNA was pH-dependent sensitively. Moreover, the helical conformation of DNA was almost retained after the intercalation and release process. These properties of the DNA/ZrRP composite suggested the potential application of layered zirconium phosphonate as a non-viral vector in gene delivery. - Graphical abstract: The intercalation of DNA into zirconium phosphonate and the release of DNA from the interlayer of zirconium phosphonate. - Highlights: ●A layered aminoethoxy-functionalized zirconium phosphonate has been synthesized. ●DNA was intercalated directly into the prepared zirconium phosphonate. ●A novel zirconium phosphonate/DNA binary hybrid was fabricated. ●DNA can be reversibly released from the interlayer of zirconium phosphonate. ●The intercalation/release processes do not induce the denaturalization of DNA.
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
Concept Feasibility Report for Electroplating Zirconium onto Uranium Foil - Year 2
The Fuel Fabrication Capability within the U.S. High Performance Research Reactor Conversion Program is funded through the National Nuclear Security Administration (NNSA) NA-26 (Office of Material Management and Minimization). An investigation was commissioned to determine the feasibility of using electroplating techniques to apply a coating of zirconium onto depleted uranium/molybdenum alloy (U-10Mo). Electroplating would provide an alternative method to the existing process of hot roll-bonding zirconium foil onto the U-10Mo fuel foil during the fabrication of fuel elements for high-performance research reactors. The objective of this research was to develop a reproducible and scalable plating process that will produce a uniform, 25 μm thick zirconium metal coating on U-10Mo foil. In previous work, Pacific Northwest National Laboratory (PNNL) established a molten salt electroplating apparatus and protocol to plate zirconium metal onto molybdenum foil (Coffey 2015). During this second year of the research, PNNL furthered this work by moving to the U-10Mo alloy system (90 percent uranium:10 percent molybdenum). The original plating apparatus was disassembled and re-assembled in a laboratory capable of handling low-level radioactive materials. Initially, the work followed the previous year’s approach, and the salt bath composition was targeted at the eutectic composition (LiF:NaF:ZrF4 = 26:37:37 mol%). Early results indicated that the formation of uranium fluoride compounds would be problematic. Other salt bath compositions were investigated in order to eliminate the uranium fluoride production (LiF:NaF = 61:39 mol% and LiF:NaF:KF = 46.5:11.5:42 mol% ). Zirconium metal was used as the crucible for the molten salt. Three plating methods were used—isopotential, galvano static, and pulsed plating. The molten salt method for zirconium metal application provided high-quality plating on molybdenum in PNNL’s previous work. A key advantage of this approach is that plating can be performed under conditions that would greatly reduce the quantity of intermetallics that form at the interface between the zirconium and U-10Mo; unlike roll bonding, the molten salt plating approach would allow for complete coverage of the U-10Mo foil with zirconium. When utilizing the experimental parameters developed for zirconium plating onto molybdenum, a uranium fluoride reaction product was formed at the Zr/U-10Mo interface. By controlling the initial plating potential, the uranium fluoride could be prevented; however, the targeted zirconium thickness (25 ±12.5 μm) could not be achieved while maintaining 100% coverage.
36 MATERIALS SCIENCE↗
Stability of zirconium hydride
This reports complete Milestone M2.2.1 Stability of zirconium hydride: Experimental investigation of the hydrogen concentration of encapsulated zirconium hydride before and after joining. In this milestone, we first report the successful fabrication of crack-free bulk zirconium hydride by using the ORNL static hydriding system. The fabricated hydride was characterized by using X-ray diffraction (XRD) and X-ray computed tomography (XCT). XRD results showed that the δ-phase zirconium hydride, α-zirconium, and ε-phase hydride were present. XCT confirmed the absence of the internal cracking. The hydrogen concentration was determined by the weight change before and after hydrogen loading, given the fact that the oxide phase was not found during XRD analysis. Two metallic cladding materials, i.e., molybdenum and FeCrAl, were selected to encapsulate the fabricated zirconium hydride pellets. Demonstration capsules were designed and machined. Mo crucible was successfully sealed by using a combination of electron beam welding and gas tungsten arc welding. Due to the undesired neutronics property of Mo, further investigation of this cladding material was not pursued. Electron beam welding and laser beam welding were employed to seal the FeCrAl capsules. XRD were performed on the zirconium hydride before and after the sealing. The results showed that the welding process has negligible impact on the hydrogen content in the FeCrAl-canned zirconium hydride. The presented work fulfilled the proposed work under this milestone and paves a pathway to fabricate single-phase zirconium hydride and to develop robust cladding materials to can bulk zirconium hydride.
36 MATERIALS SCIENCE↗
Investigation and modeling of the degradation of zirconium-based fuel cladding during corrosion in steam and air-steam mixtures at high temperatures - 2016-0033
This paper presents the main results of a study conducted to quantify and to model the degradation state of zirconium-based fuel claddings submitted to severe-accident conditions in a nuclear reactor core: high temperatures and either pure steam or an air-steam mixture. Due to the progressive thickening of a dense and protective ZrO{sub 2} layer, the oxidation kinetics of zirconium-based claddings in steam at high temperatures typical of severe nuclear accidents are generally cubic or parabolic. However, for some temperature domains, this oxide layer may crack, becoming porous and no longer protective. In these 'breakaway' conditions, the oxidation kinetics change from (sub)parabolic to linear or even accelerated. In addition, the temperature increase can lead core materials to melt and to relocate down to the vessel lower head, threatening its integrity. If it fails, and for specific conditions, air ingress may take place into the reactor. Hence, oxygen and nitrogen both react with zirconium-based claddings successively through the oxidation of zirconium (forming a ZrO{sub 2} layer), nitriding of zirconium (forming zirconium nitride particles), and the oxidation of zirconium nitride (forming ZrO{sub 2} and releasing nitrogen). These self-sustained chemical reactions enhance the deterioration of zirconium-based claddings and their ZrO{sub 2} layers, inducing an increase in their open porosity. To quantify this porosity, a series of two-step experiments was conducted. First, ZIRLO{sup TM} cladding samples were isothermally oxidized in pure steam or in a 50:50 mol% air-steam mixture at several different temperatures and durations. The main thermal effects on reaction kinetics and the high impact of air on the cladding degradation were all confirmed by experimental results. Second, pioneering porosimetry measurements by mercury intrusion were realized for the first time on such corroded cladding samples. In both atmospheres, it was pointed out that 1,200 and 1,250 K led to particularly porous oxide layers, especially due to strong breakaway effects. Moreover, we confirmed that the presence of air strongly enhances the oxide cracking: cladding samples were more porous when oxidized in the air-steam mixture than under pure steam. Finally, we observed that in all conditions, the open porous volume fraction of ZIRLO claddings continuously rose during their corrosion process. Hence, for each experimental condition, porosity correlations were determined through linear regressions, and porosity increase rates were deduced by derivation versus time and validated against porosimetry results of cladding samples corroded in transient nonisothermal) conditions. (authors)
36 MATERIALS SCIENCE↗
Separation of Zirconium from Zircaloy-2 for Spent Nuclear Fuel Cladding Applications using a Chloride Volatility Process
The United States produces a large amount of cladding material from used nuclear fuel, expected to approach 1,000 MT/year for the next 50 years. The recovery of zirconium from the cladding material would represent a ∼50% savings on the geologic waste disposal requirement, and savings of >$40 M/year due to the cost of hafnium-free zirconium. Previous simulation studies have shown that in the irradiated cladding, the major activation products found are {sup 125}Sb, {sup 60}Co, and {sup 93m}Nb, and the major fission product found is {sup 137}Cs. These radioisotopes make it impossible to treat the cladding as low level waste. In a chlorination method, zirconium cladding reacts with Cl{sub 2} at 350-380 deg. C to produce mainly ZrCl{sub 4}. The zirconium tetrachloride has a relatively low boiling point (sublimes at 330 deg. C) which allows its separation from other products, and so it can be treated as low level waste for disposal or recycling. This chlorination approach has been proposed for the reprocessing of zirconium. However, previous work has observed element impurities (Fe, Sn, Cr, Cs, Sb, and Nb) in the recovered ZrCl{sub 4} meaning that additional steps are necessary in a chlorination process previous to the reduction of ZrCl{sub 4} to the metallic form. The chlorination process for zirconium reprocessing has been highly encouraged as it has proven more efficient than previous mechanical processes, redox processes, and other chemical methods such as iodination. The overall goal of this project is to develop the road map for a chlorination process that will yield high purity zirconium. The current study focuses on the recovery of ZrCl{sub 4} from non-irradiated alloys used as nuclear fuel cladding. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
First-principles study of mechanical properties of zirconium alloys and hydrides
Zirconium alloys have been extensively used as fuel cladding materials in nuclear light water reactors (LWRs) due to their favorable mechanical properties, corrosion resistance, low thermal neutron capture cross section and criticality. Because the main functions of these structural materials are to maintain the integrity of UO{sub 2} fuel rods and prevent the release of fission products into the biosphere, a detailed knowledge of their structural stability is of critical importance to ensure the safe operation of nuclear reactors, and the safety of storage, transportation or disposition of used nuclear fuel assemblies. While the Zr-H phase diagram has been known since the 1950's and has remained largely unchanged, the structures, formation mechanisms, and stability of Zr-hydride phases in pure α-Zr or Zr alloys matrices have been subjects of much debate over the last few decades. The formation of hydride phases in the Zr matrix has been linked in previous studies to the degradation of mechanical properties of Zr/Zr-alloys materials. Therefore, an accurate knowledge of the interplay between the structures, stability and mechanical properties of Zr/Zr-alloys and Zr hydrides is of paramount significance to ensure materials reliability during their in- service lifetime. Despite few first-principles studies have investigated the mechanical stability of Zr hydrides polymorphs proposed so far, no ab initio studies have been carried out, to the best of our knowledge, on the mechanical properties of modern nuclear-grade zirconium alloys such as Zircaloy-4 (Zry-4; 1.2-1.7% Sn), Zirconium low oxidation alloy (ZIRLO; 0.7-1.0% Sn and ∼1.0% Nb), or M5 alloy (0.8-1.2% Nb). Therefore, it is important to carry out a series of first-principles calculations systematically to investigate the structure-property relationship of zirconium hydrides and zirconium alloys, that can help explain differences in elastic properties between high-purity zirconium and nuclear-grade zirconium alloys. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Alloying effect of niobium and tin on the zirconium alloy fuel cladding behavior at high temperature oxidation in steam - 2016-0072
As the materials for fuel claddings for water-cooled reactors, binary and multicomponent zirconium alloys zirconium-niobium (Zr-Nb) (E110, M5), zirconium-tin-iron (Zr-Sn-Fe) (Zircaloy-2,-4), and zirconium-niobium-tin-iron (Zr-Nb-Sn-Fe) (E635, ZIRLO{sup R} ) most commonly are used. Improvement extends, in particular, by varying their composition by niobium, tin, and iron. At the same time, importance is given to ensuring the safe operation of the fuel rods not only in normal conditions but also in emergency situations such as a loss of coolant accident. In this paper, we present the research results of studies of the influence of alloying elements niobium, tin, and iron on corrosion resistance and embrittlement at high-temperature steam oxidation of fuel claddings based on the alloy types Zr-xNb (x=1.0/2.5) and Zr-xNb-ySn-zFe (x=0.6/2.4; y=0.24/1.1; z=0.18/0.34) made by using zirconium sponge. We conducted high-temperature steam oxidation tests at temperatures of 1,000 deg. C, 1,100 deg. C, and 1,200 deg. C with continuous measurement of the weight gain during the experiment and subsequent cooling in steam with rate ∼ 20 deg. C/s. We conducted studies of kinetics of high-temperature oxidation in steam, structural-phase state changes, alloying elements distribution, absorbed hydrogen content, and residual ductility after rapid cooling of the oxidized specimens. We revealed the difference in the oxidation kinetics of the materials studied, which decreased as the oxidation temperature increased. Likewise, an increase of the tin content in the alloy influenced a specimen's breakaway oxidation intensification at 1,000 deg. C. This shows that the mechanical properties of fuel claddings oxidized under the same conditions depend on their alloying composition, and the hydrogen fraction is absorbed by the specimen and formed by the oxidation of ZrO{sub 2}, α-Zr(O), and ex-β layer structures, which in turn varies depending on the alloy composition. The increase of niobium, tin, and iron content in the zirconium alloy leads to a decrease in the residual ductility of fuel claddings after high-temperature steam oxidation. According to obtained results, among the materials studied, Zr-1.0Nb alloy is the most resistant to embrittlement during high-temperature steam oxidation at temperatures of 1,000-1,200 deg. C. (authors)
36 MATERIALS SCIENCE↗
Diffusion bonding of tungsten-vanadium-zirconium using vacuum hot pressing for the development of a low decay heat cladding solution for tungsten spallation targets
Tantalum has been used as cladding material for water-cooled solid tungsten targets at many leading spallation neutron production facilities thanks to its high neutron yield, manageable radiation damage behavior, and excellent corrosion/erosion resistance in radiation environments. However, from a safety hazard perspective, thermal neutron capture of tantalum in spallation environments causes a high specific decay heat in the target volume, which often becomes a limiting factor in increasing the beam power on the target. In this paper, we studied vacuum hot pressing (VHP) parameters to diffusion bond zirconium to tungsten to explore the feasibility of using zirconium alloys as an alternative cladding material to tantalum. Zirconium alloys have long been used as cladding material for early generation solid spallation targets, and nuclear fuel rods. In spallation environments zirconium has significantly lower decay heat with shorter decay time compared to tantalum. The hot isostatic pressing (HIP) of zirconium and tungsten is known to produce limited bonding quality due to the formation of the brittle ZrW2 intermetallic layer. To overcome this problem, placing a vanadium interlayer between tungsten and zirconium has been proposed by exploring parameter space in binary alloy phase diagrams. Under the VHP conditions, 860 ◦C at 70 MPa for 4 hours, Zr-V and V-W showed good diffusion bonding, which demonstrates the feasibility of a single step HIP process to make the zirconium alloy clad tungsten spallation volumes.
Lee, Yong Joong↗
Capture of Tritium Released from Cladding in the Zirconium Recycle Process
Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-based cladding and could be released from the cladding when the solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using nonradioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Capture of Tritium Released from Cladding in the Zirconium Recycle Process
This report is issued as the first revision to FCRD-MRWFD-2016-000297. Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-based cladding and could be released from the cladding when the solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using non-radioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
The Effect of Boron and Zirconium on the Structure and Tensile Properties of the Cast Nickel-Based Superalloy ATI 718Plus
The effects of boron and zirconium on cast structure, hardness, and tensile properties of the nickel-based superalloy 718Plus were investigated. For this purpose, five alloys with different contents of boron and zirconium were cast via vacuum induction melting and then purified via vacuum arc remelting. Microstructural analysis by light-optical microscope and scanning electron microscope equipped with energy-dispersive x-ray spectroscopy and phase studies by x-ray diffraction analysis were performed. The results showed that boron and zirconium tend to significantly reduce dendritic arm spacing and increase the amount of Laves, Laves/gamma eutectic, and carbide phases. It was also found that boron led to the formation of B{sub 4}C and (Cr, Fe, Mo, Ni, Ti){sub 3}B{sub 2} phases and zirconium led to the formation of intermetallic phases and ZrC carbide. In the presence of boron and zirconium, the hardness and its difference between dendritic branches and inter-dendritic spaces increased by concentrating such phases as Laves in the inter-dendritic spaces. These elements had a negative effect on tensile properties of the alloy, including ductility and strength, mainly because of the increase in the Laves phase. It should be noted that the largest degradation of the tensile properties occurred in the alloys containing the maximum amount of zirconium.
36 MATERIALS SCIENCE↗
Anticorrosive Behavior and Porosity of Tricationic Phosphate and Zirconium Conversion Coating on Galvanized Steel
This work evaluates the corrosion resistance of galvanized steel treated with tricationic phosphate and zirconium conversion coating after painting, by using electrochemical techniques, accelerated and field corrosion tests. A non-uniform and heterogeneous distribution of zirconium on the steel surface was observed due to preferential nucleation of the zirconium on the aluminum-rich sites on the surface of galvanized steel. The long-term anti-corrosion performance in a saline solution was better for the phosphate coating up to 120 days. The coating capacitance registered a higher increase for the zirconium coatings than the phosphate coatings up to 120 days of immersion. This result agrees with the higher porosity of zirconium coating in relation to the phosphate coating. After 3840 h of accelerated corrosion test, and after 1 year of accelerated field test, zirconium-treated samples showed an average scribe delamination length higher than the phosphate-treated samples.
36 MATERIALS SCIENCE↗
Nanoindentation study of bulk zirconium hydrides at elevated temperatures
Here, the mechanical properties of zirconium hydrides was studied using nano-indentation technique at a temperature range of 25 – 400 °C. Temperature dependency of reduced elastic modulus and hardness of δ- and ε-zirconium hydrides were obtained by conducting nanoindentation experiments on the bulk hydride samples with independently heating capability of indenter and heating stage. The reduced elastic modulus of δ-zirconium hydride (H/Zr ratio =1.61) decreased from ~113 GPa to ~109 GPa while temperature increased from room temperature to 400°C. For ε-zirconium hydrides (H/Zr ratio=1.79), the reduced elastic modulus decreased from 61 GPa to 54 GPa as temperature increased from room temperature to 300 °C. Whereas, hardness of δ-zirconium hydride significantly decreased from 4.1 GPa to 2.41 GPa when temperature increased from room temperature to 400 °C. Similarly, hardness of ε-zirconium hydride decreased from 3.06 GPa to 2.19 GPa with temperature increase from room temperature to 300°C.
36 MATERIALS SCIENCE↗
Modification in band gap of zirconium complexes
The optical properties of zirconium complexes with amino acid based Schiff bases are reported here. The zirconium complexes show interesting stereo chemical features, which are applicable in organometallic and organic synthesis as well as in catalysis. The band gaps of both Schiff bases and zirconium complexes were obtained by UV-Visible spectroscopy. It was found that the band gap of zirconium complexes has been modified after adding zirconium compound to the Schiff bases.
75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗