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At least 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Methods to Capture the Tritium Released by Recovery/Recycle of Zirconium from Used Fuel Cladding: A Review

After uranium, the second largest mass contained in most used nuclear fuel (UNF) is the zirconium-based Zircaloy{sup R} cladding in the fuel cladding which accounts for about 25% of the mass. In current practice, the cladding requires disposal in a geologic repository. Zirconium may be recovered from the Zircaloy cladding of UNF for recycle. Alternatively, it may be decontaminated to recategorize it as low-level waste. Either approach reduces the quantities of high-level waste destined for a geologic repository, thereby reducing the overall cost of recycling UNF. Process development studies are under way in the United States and Republic of Korea based on dry chlorination using a mixture of chlorine (Cl{sub 2}) and an inert carrier such as argon (Ar) to convert the zirconium in the alloy, which is greater than 98 wt% of the alloy mass, to zirconium tetrachloride (ZrCl{sub 4}). The ZrCl{sub 4} is volatile and is easily separated from the non-volatile components. The residues include non-volatile chlorides such as CrCl{sub 3}, oxychlorides and oxides. The product ZrCl{sub 4} is recovered as a solid salt by condensation when cooled to less than about 300 deg. C. Feasibility tests with actual UNF cladding have demonstrated that the recovered zirconium chloride salt is sufficiently separated from fuel and fission product radionuclides that it emanates no significant dose and can be handled without shielding. Several methods for capturing tritium released from cladding during the chlorination process have been identified. A detailed description of the methodologies, advantages, drawbacks, and conceptual flowsheets will be presented. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Investigations into the Gas Phase Reactions of Zirconium Tetrachloride (ZrCl) from UNF Cladding

Zirconium-based cladding materials that surround used nuclear fuel (UNF) rods represent a significant fraction of the total amount of high level waste that must be disposed of in long term geological repositories. The decontamination, recovery and reuse of these materials would significantly reduce the high level waste associated with light water reactors and could provide a new source of hafnium free zirconium for the fabrication of new nuclear components. A step in a strategy to recycle zirconium from used cladding materials is the high yield chlorination of Zircaloys to produce zirconium tetrachloride (ZrCl{sub 4}). The ZrCl, is contaminated with other metal chloride species that originate from the alloying metals as well as contaminants from oxide layers on the cladding, trace fission products, fuel residues, activation byproducts, and tritium. The properties of these complex mixtures must be better understood before purification protocols can be developed that will allow zirconium chloride to be recycled further. First order thermodynamic modeling of high temperature mixtures of ZrCl, and other metal chlorides in the gas phase may not accurately predict their properties. Computational modeling studies indicate that mixed metal adducts may form under the conditions of sublimation-based purification protocols. In light of these predictions, alternate strategies may also be needed to obtain pure ZrCl, from such mixtures. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Overcoming the crystallization and designability issues in the ultrastable zirconium phosphonate framework system

Metal-organic frameworks (MOFs) based on zirconium phosphonates exhibit superior chemical stability suitable for applications under harsh conditions. These compounds mostly exist as poorly crystallized precipitates, and precise structural information has therefore remained elusive. Furthermore, a zero-dimensional zirconium phosphonate cluster acting as secondary building unit has been lacking, leading to poor designability in this system. We overcome these challenges and obtain single crystals of three zirconium phosphonates that are suitable for structural analysis. Furthermore, these compounds are built by previously unknown isolated zirconium phosphonate clusters and exhibit combined high porosity and ultrastability even in fuming acids. SZ-2 possesses the largest void volume recorded in zirconium phosphonates and SZ-3 represents the most porous crystalline zirconium phosphonate and the only porous MOF material reported to survive in aqua regia. SZ-2 and SZ-3 can effectively remove uranyl ions from aqueous solutions over a wide pH range, and we have elucidated the removal mechanism.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2018

Design of Laboratory System to Recover Tritium from Zirconium Recycle and Plan to Test System

The objective of the tritium separations task is to develop a viable process to separate tritium from the zirconium recycle off-gas stream. In support of this long-term objective, the goal of the present work is to design a test system and prepare a test plan for an alternate tritium capture method suitable for, and compatible with, recycle of the reactive gas (i.e., chlorine) in the zirconium recycle process. The zirconium recycle process is under development to recover zirconium from the Zircaloy® cladding component of used nuclear fuel. Small quantities of all isotopes of hydrogen (all H) are trapped in the Zircaloy as zirconium hydride. Processing releases and converts this hydrogen to hydrogen chloride, including tritium-substitu ted hydrogen chloride, 3 HCl (a.k.a., TCl). This compound of tritium must be removed from the zirconium recycle process off-gas and sequestered to satisfy regulations governing emissions of tritium.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Thermomechanical Processing of Uranium Alloys with 10 and 50 Weight Percent Zirconium

This work focuses on casting and thermomechanical processing for moderate sized (500-1000g) 10% by weight and 50% by weight uranium-zirconium alloys. Three castings of uranium – 10wt% zirconium and a fourth casting of uranium – 50wt% zirconium were completed using different castings processes. The first casting process relied on only induction mixing within the VIM system thereby eliminating the need for an arc-melting feedstock preparation step. The second casting process used the arc-melting process prior for feedstock to the same VIM system. In this study, the induction mixing process alone with a graphite crucible appears insufficient for the fabrication of UZr alloys. Graphite is suitable with the use of master alloy UZr feedstock produced by arc-melting. There was a moderate increase of approximately 100 ppm of carbon within the U10Zr alloys. However, there was a near 1000 ppm increase of carbon in the U50Zr alloy. This carbon content is considered too high for forming processes, and therefore, warrants the use of a ceramic crucible for UZr castings near 50wt% zirconium. Zirconia does not appear suitable for UZr castings near 50wt% zirconium. The use of Beryllia or Thoria may be a more suitable option. However, due to the significantly lower liquidus and lower zirconium content, Zirconia crucibles do appear viable for carbon reduction in the U10Zr system. Two thermomechanical processes are also studied in this work - hot extrusion of the U50Zr system and heat treatment in an argon atmosphere. This work demonstrated that extrusion of the U50Zr alloy near 680°C is a viable choice even with +1000 ppm carbon. Extrusion of the U50Zr system near 630°C was unsuccessful and does not appear viable unless significant force and tooling strength is available. There appears to be a moderate increase in extrudate hardness when compared to as-cast. Heat-treatment reduced the hardness of as-cast alloys in the U10Zr castings as well as resolve some variance in the microstructures. However, for the U50Zr billet that did not successfully extrude at 630°C, heat treatment at 900°C for 96 hours did not have a significant impact on hardness or microstructure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Technology Enabling Zero-EPZ Micro Modular Reactors: Fabrication of Zirconium Hydride with Controlled Hydrogen Loading

This report completes Milestone M2.2.2 - ZrH pellet fabricated with controlled hydride loading: ZrH1.5±0.1 single phase pellets fabricated with controlled and thermally stable hydride loading. Hydrogen content to be quantified and single-phase confirmed by X-ray diffraction. In this milestone, we report the successful fabrication of delta-phase zirconium hydride pellets using the ORNL bulk metal hydriding system. The successful deployment of zirconium hydride moderator in advanced reactors requires development of a consistent and affordable production pathway along with implementation of a hydrogen retention solution throughout the reactor life. Fabrication of delta-phase zirconium hydride is challenging since the absorption of a large amount of hydrogen into alpha-zirconium induces significant volume expansion and the present hydrogen concentration gradient results in cracking. A fully programmable hydriding system with continuous hydrogen partial pressure and flow control to facilitate processing of massive metal hydride has been developed at ORNL. In this report, the working principle of the hydriding system will be introduced. Characterization of the produced zirconium hydride includes X-ray powder diffraction to identify the present phases and X-ray Computed Tomography to visualize the internal microstructure (including cracks). The results indicate that single delta-phase zirconium hydride pellets with various sizes have been successfully produced.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Methods to increase solid solution zirconium in aluminum alloys

A method of making an aluminum alloy containing zirconium includes heating a first composition comprising aluminum to a first temperature of greater than or equal to about 580° C. to less than or equal to about 800° C. The method further includes adding a second composition including a copper-zirconium compound to the first composition to form a third composition. The copper-zirconium compound of the second composition has a molar composition of greater than or equal to about 41% zirconium to less than or equal to about 67% zirconium and a balance of copper. The method also includes solidifying the third composition at a cooling rate of greater than or equal to about 0.1° C./second to less than or equal to about 100° C./second to a second temperature less than or equal to a solidus temperature and decomposing the copper-zirconium compound at a third temperature of less than or equal to about 715° C.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Preliminary irradiation effect on corrosion resistance of zirconium alloys - 2016-0067

In order to improve the composition and structure of zirconium alloys, it is important to know how changes in their structure-phase state under irradiation affect various properties, including corrosion. In this work, autoclave corrosion tests were carried out on fuel cladding samples made of zirconium-niobium and zirconium-niobium-tin-iron alloy systems. All samples were preliminarily irradiated in a BOR-60 nuclear reactor at a fluence of 2 X10{sup 26} m{sup -2} (E>0.1 MeV). The autoclave tests of irradiated and unirradiated samples were performed at 350 deg. C for 240 days in distilled water containing 10 ppm lithium and 1,600 ppm boron. The irradiation of all alloys resulted in a significant change in their initial structure-phase state. The content of niobium in b-niobium-phase precipitates was reduced from 80 % to - 90 % to 50 % to - 60 %. Iron atoms from the Laves-phase particles were diffused into a matrix, and their lattice transformed from hexagonal close-packed to body-centered cubic. An average size of the transformed second-phase precipitates in the irradiated alloys increased relatively to an average size of the initial precipitates. Elongated radiation-induced precipitates (RIPs) with an average length up to 7 nm were formed in all alloys under irradiation. In comparison with the unirradiated state, irradiation reduced the corrosion of zirconium-niobium alloys, and this effect became more apparent with increased niobium content in the alloy. For zirconium-niobium-tin-iron alloys, irradiation led to more severe corrosion, mostly because of increasing tin content. Niobium content increased up to 2.37 %, and iron content decreased up to 0.18 % in the alloys, contributing to higher corrosion resistance. An interconnection between the niobium/(tin + iron) alloying parameter, quantity of nanometric precipitates formed therein under irradiation, and their corrosion performance has been suggested. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2014

Composition and microstructure of zirconium and hafnium germanates obtained by different chemical routes

The phase composition and morphology of zirconium and hafnium germanates synthesized by ceramic and co-precipitation routes were studied. The products were characterized using high-temperature X-ray diffraction analysis (XRD), Raman spectroscopy, scanning electron microscopy (SEM) and thermal (TG/DTA) analysis. To investigate the phase composition and stoichiometry of compounds the unit cell parameters were refined by full-profile Rietveld XRD analysis. The morphology of products and its evolution during high-temperature treatment was examined by SEM analysis. It was stated that there is the strong dependence of the phase composition and morphology of products on the preparation route. The ceramic route requires a multi-stage high-temperature treatment to obtain zirconium and hafnium germanates of 95% purity or more. Also, there are strong diffusion limitations to obtain hafnium germanate Hf{sub 3}GeO{sub 8} by ceramic route. On the contrary, the co-precipitation route leads to the formation of nanocrystalline single phase germanates of stoichiometric composition at a relatively low temperatures (less than 1000 °C). The results of quantitative XRD analysis showed the hafnium germanates are stoichiometric compounds in contrast to zirconium germanates that form a set of solid solutions. This distinction may be related to the difference in the ion radii of Zr and Hf. - Graphical abstract: The phase composition and morphology of zirconium and hafnium germanates synthesized by ceramic and co-precipitation routes were studied. It was stated that there is the strong dependence of the phase composition and morphology of products on the preparation route. Display Omitted - Highlights: • Zr and Hf germanates were synthesized by ceramic and co-precipitation routes. • The morphology of products depends on the synthesis parameters. • Zirconium germanates forms a set of solid solutions. • Hafnium germanates are stoichiometric compounds.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2013

Crystal-structure refinement of zirconium-rich eudialyte and its place among calcium-poor eudialyte-group minerals

The repeated refinement of the crystal structure of zirconium-rich eudialyte based on the X-ray diffraction data set collected earlier revealed new structural features. The trigonal unit-cell parameters are a = 14.222(3) A, c = 30.165(5) A, V = 5283.9 A{sup 3}. The refinement resulted in the reduction of the R factor from 0.045 (2347F > 4{sigma}(F)) to 0.035 (3124F > 3{sigma}(F)). It was found that the ordering of Ca and Fe in six-membered rings leads to the lowering of the symmetry to R3. An excess amount of zirconium (more than three atoms per symmetrically independent unit) is located in the M2 microregion in square and five-vertex polyhedral positions. However, this amount is insufficient to be dominant, and the deficiency of zirconium is compensated for by sodium atoms. Based on the new data, zirconium-rich eudialyte can be assigned to the oneillite subtype, being a zirconium-rich and aluminum variety of raslakite.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Understanding irradiation growth through atomistic simulations: Defect diffusion and clustering in α-zirconium and the influence of alloying elements - 2016-0093

Irradiation-induced structural changes of α-zirconium alloys and in particular the effect of iron were investigated by molecular dynamics simulations using embedded atom potentials derived from first-principles calculations. The simulations revealed that at temperatures between 500 and 600 K self-interstitial atoms (SIAs) diffuse rapidly in a cooperative movement, preferably parallel to basal planes (a directions; ), forming nano-clusters with an extension in and . Vacancies diffuse more slowly than SIAs and remain isolated for a longer period of time. Nano-clusters associated with SIAs cause a pronounced overall expansion in a directions, as well as local strains. Under compressive strain in the c direction, vacancy diffusivity increases in the c direction. In contrast, the diffusivity of SIAs increases in the c direction under a tensile strain in the c direction. SIA nano-clusters are highly mobile within basal planes. Vacancy clusters grow by merging, leading to a contraction in the a direction, compensating for the expansion caused by SIA nano-clusters and possibly contributing to the plateau in growth after the initial rapid expansion. At the onset of breakaway growth, possibly due to stress buildup, the vacancy nano-clusters can condense into c loops, thereby diminishing the compensation effect. The alloying elements iron, nickel, chromium, and niobium liberated from secondary phase particles under irradiation or already in solution are attracted to vacancies and SIAs and are found inside vacancy and SIA loops. The interaction of alloying elements with defect clusters is discussed, with a particular focus on iron. Iron has been found to promote cluster formation in zirconium, and the structures of zirconium-iron clusters have been analyzed. Tin is repelled by SIA clusters and only weakly attracted by vacancies. Niobium impedes the diffusion of SIAs (and therefore may increase annihilation rates with nearby vacancies) and does not destabilize vacancy or SIA clusters. Ab initio calculations of the dimensional and elastic coefficients of the intermetallic phases occurring in secondary phase particles, such as Zr{sub 2}Fe and Zr{sub 3}Fe, are presented, allowing an assessment of local strains in a zirconium matrix. Thus, novel results from extended molecular dynamics simulations provide new insights and contribute to a deeper understanding of the complex mechanisms causing irradiation-induced dimensional changes and the breakaway growth of zirconium alloys. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2015

White Paper Summary of 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding

This white paper recommends that ASTM International develop standards to address the potential impact of hydrides on the long term performance of irradiated zirconium alloys. The need for such standards was apparent during the 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding and Assembly Components, sponsored by ASTM International Committee C26.13 and held on June 10-12, 2014, in Jackson, Wyoming. The potentially adverse impacts of hydrogen and hydrides on the long term performance of irradiated zirconium-alloy cladding on used fuel were shown to depend on multiple factors such as alloy chemistry and processing, irradiation and post irradiation history, residual and applied stresses and stress states, and the service environment. These factors determine the hydrogen content and hydride morphology in the alloy, which, in turn, influence the response of the alloy to the thermo-mechanical conditions imposed (and anticipated) during storage, transport and disposal of used nuclear fuel. Workshop presentations and discussions showed that although hydrogen/hydride induced degradation of zirconium alloys may be of concern, the potential for occurrence and the extent of anticipated degradation vary throughout the nuclear industry because of the variations in hydrogen content, hydride morphology, alloy chemistry and irradiation conditions. The tools and techniques used to characterize hydrides and hydride morphologies and their impacts on material performance also vary. Such variations make site-to-site comparisons of test results and observations difficult. There is no consensus that a single material or system characteristic (e.g., reactor type, burnup, hydrogen content, end-of life stress, alloy type, drying temperature, etc.) is an effective predictor of material response during long term storage or of performance after long term storage. Multi-variable correlations made for one alloy may not represent the behavior of another alloy exposed to identical conditions and the material responses to thermo-mechanical exposures will be different depending on the materials and systems used. The discussions at the workshop showed several gaps in the standardization of processes and techniques necessary to assess the long term performance of irradiated zirconium alloy cladding during dry storage and transport. The development of, and adherence to, standards to help bridge these gaps will strengthen the technical basis for long term storage and post-storage operations, provide consistency across the nuclear industry, maximize the value of most observations, and enhance the understanding of behavioral differences among alloys. The need for, and potential benefits of, developing the recommended standards are illustrated in the various sections of this report.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Chemical Vapor Deposition of Zirconium Compounds: A Review

Coatings of zirconium compounds are used in a wide variety of fields, yet an understanding and descriptions of deposition mechanisms are scant in the public literature. The mechanisms of deposition for metallic zirconium, ZrC, ZrN, ZrO 2 , ZrB 2 , and zirconium silicides are discussed based on the direct vapor deposition research of those compounds where possible or compared to complementary titanium systems when direct research is lacking. Both inorganic and organometallic deposition systems are discussed. As a class of compounds, an understanding of the vapor deposition mechanisms can be significantly improved by investigations on metallic zirconium deposition by zirconium halides and hydrogen and by in situ analysis techniques such as Fourier-transform infrared (FTIR) spectroscopy or x-ray photoelectron spectroscopy (XPS).

organometallic↗
DOE OSTI2017

Purification of Zirconium Tetrachloride (ZrCl{sub 4}) from UNF Cladding - A Progress Report

Zirconium-based cladding materials that surround used nuclear fuel (UNF) rods represent a significant fraction of the total amount of high level waste that must be disposed of in long term geological repositories. The decontamination and/or recovery and reuse of these materials would significantly reduce the amount of high level waste associated with light water nuclear energy reactors and could provide a new source of hafnium free zirconium for the fabrication of new nuclear components. A step in a strategy to recycle zirconium from used cladding materials is the direct, high yield chlorination reaction that produces zirconium tetrachloride (ZrCl{sub 4}). The ZrCl{sub 4} so obtained is contaminated with other metal chloride species (FeCl{sub 3}, NbCl{sub 5}, NbOCl{sub 3}, SnCl{sub 4}, CsCl, SbCl{sub 5}, etc.) that originate from the alloying metals present initially as well as contaminants from oxide layers on the cladding, trace fission products, fuel residues, activation byproducts, and tritium. The suggested strategy to remove the metal chloride impurities is based on sublimation at atmospheric pressure combined with chemical treatments influencing volatility of the impurities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Alternate Methods for Cleaning Zirconium Plate

The United States High Performance Research Reactor Conversion Program identified the need to transition from highly enriched uranium fuel to lowly enriched uranium fuel for high power research and civil reactors. One of the fuel configurations selected for these reactors was a uranium (U) foil alloyed with 10% molybdenum (Mo). The U-10Mo fuel foil was then covered with a zirconium (Zr) interlayer and pressed into an aluminum cladding. Idaho National Laboratory prescribed a need to remove surface oxidation on the Zr prior to bonding it with U-10Mo fuel foil. To determine the best method of oxidation removal, a variety of polishing methods, including hand polishing with a diamond paste or Scotch-Brite™ pads, mechanically wet-polishing with abrasive belts, and chemical etching with hydrofluoric/nitric acid mixtures, were investigated on pre-polished Zircaloy 702 sheets purchased directly from the manufacturer. No change in surface roughness was detected using any of these methods relative to the as-received material. In fact, scanning electron microscopy detected grit from hand polishing with Amplex Grade 30 waste soluble diamond paste and Scotch-Brite™ pads. Visible staining was found, and fluorine detected via X-ray photoelectron spectroscopy on the flash etching samples. These contaminants were not seen on the as-received material. In addition to the need to remove the oxide layer on the Zr prior to co-rolling, there are concerns that the Neolube debonding agent on the co-rolling can may contaminate the Zr surface after co-rolling. The experimental results found that zirconium in contact with Neolube coated steel plates was easily cleaning with ethanol wiping after heat treated and no carbide inclusions remained on the surface. Overall, all cleaning methods tested did not significantly change the surface roughness of the plates or reduce the oxide layer present on the zirconium. Carbon was found on all the samples, before and after cleaning, except for the flash etch samples which detected fluorine instead. The insignificant changes seen in the cleaned plates relative to the as received plates question the need for cleaning prior to rolling. Further investigation in the requirements for cleaning at this step is recommended. Current methods for cleaning the surface after hot rolling with alcohol appear sufficient. After hot rolling, if Neolube does peel off the can and onto the zirconium coated foil surface, a dry abrasive or diamond paste polish is not recommended due to the chance of abrasive being embedded into the surface. However, the ability of an etchant to lift the lubricant off the zirconium surface indicates that a spot etch method could work to clean specific spots where Neolube is present, if the need arises.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1972

A simple spectrophotometric method for determination of zirconium or hafnium in selected molybdenum-base alloys

A simple analytical procedure is described for accurately and precisely determining the zirconium or hafnium content of molybdenum-base alloys. The procedure is based on the reaction of the reagent Arsenazo III with zirconium or hafnium in strong hydrochloric acid solution. The colored complexes of zirconium or hafnium are formed in the presence of molybdenum. Titanium or rhenium in the alloy have no adverse effect on the zirconium or hafnium complex at the following levels in the selected aliquot: Mo, 10 mg; Re, 10 mg; Ti, 1 mg. The spectrophotometric measurement of the zirconium or hafnium complex is accomplished without prior separation with a relative standard deviation of 1.3 to 2.7 percent.

Dupraw, W. A.↗
NASA NTRS1984

Zirconium fluoride glass - Surface crystals formed by reaction with water

The hydrated surfaces of a zirconium barium fluoride glass, which has potential for application in optical fibers and other optical elements, were observed by scanning electron microscopy. Crystalline zirconium fluoride was identified by analysis of X-ray diffraction patterns of the surface crystals and found to be the main constituent of the surface material. It was also found that hydrated zirconium fluorides form only in highly acidic fluoride solutions. It is possible that the zirconium fluoride crystals form directly on the glass surface as a result of its depletion of other ions. The solubility of zirconium fluoride is suggested to be probably much lower than that of barium fluoride (0.16 g/100 cu cm at 18 C). Dissolution was determined to be the predominant process in the initial stages of the reaction of the glass with water. Penetration of water into the glass has little effect.

Doremus, R. H.↗
DOE OSTI2014

Zirconium-based alloys, nuclear fuel rods and nuclear reactors including such alloys, and related methods

Zirconium-based metal alloy compositions comprise zirconium, a first additive in which the permeability of hydrogen decreases with increasing temperatures at least over a temperature range extending from 350.degree. C. to 750.degree. C., and a second additive having a solubility in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. At least one of a solubility of the first additive in the second additive over the temperature range extending from 350.degree. C. to 750.degree. C. and a solubility of the second additive in the first additive over the temperature range extending from 350.degree. C. to 750.degree. C. is higher than the solubility of the second additive in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. Nuclear fuel rods include a cladding material comprising such metal alloy compositions, and nuclear reactors include such fuel rods. Methods are used to fabricate such zirconium-based metal alloy compositions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗