Modelling zirconium hydrides using the special quasirandom structure approach

KAUST Repository

Wang, Hao; Chroneos, Alexander I.; Jiang, Chao; Schwingenschlö gl, Udo

2013-01-01

The study of the structure and properties of zirconium hydrides is important for understanding the embrittlement of zirconium alloys used as cladding in light water nuclear reactors. Simulation of the defect processes is complicated due to the random distribution of the hydrogen atoms. We propose the use of the special quasirandom structure approach as a computationally efficient way to describe this random distribution. We have generated six special quasirandom structure cells based on face centered cubic and face centered tetragonal unit cells to describe ZrH2-x (x = 0.25-0.5). Using density functional theory calculations we investigate the mechanical properties, stability, and electronic structure of the alloys. © the Owner Societies 2013.

Evaluation of hydride blisters in zirconium pressure tube in CANDU reactor

International Nuclear Information System (INIS)

Cheong, Y. M.; Kim, Y. S.; Gong, U. S.; Kwon, S. C.; Kim, S. S.; Choo, K.N.

2000-09-01

When the garter springs for maintaining the gap between the pressure tube and the calandria tube are displaced in the CANDU reactor, the sagging of pressure tube results in a contact to the calandria tube. This causes a temperature difference between the inner and outer surface of the pressure tube. The hydride can be formed at the cold spot of outer surface and the volume expansion by hydride dormation causes the blistering in the zirconium alloys. An incident of pressure tube rupture due to the hydride blisters had happened in the Canadian CANDU reactor. This report describes the theoretical development and models on the formation and growth of hydride blister and some experimental results. The evaluation methodology and non-destructive testing for hydride blister in operating reactors are also described

Evaluation of hydride blisters in zirconium pressure tube in CANDU reactor

Energy Technology Data Exchange (ETDEWEB)

Cheong, Y M; Kim, Y S; Gong, U S; Kwon, S C; Kim, S S; Choo, K N

2000-09-01

When the garter springs for maintaining the gap between the pressure tube and the calandria tube are displaced in the CANDU reactor, the sagging of pressure tube results in a contact to the calandria tube. This causes a temperature difference between the inner and outer surface of the pressure tube. The hydride can be formed at the cold spot of outer surface and the volume expansion by hydride dormation causes the blistering in the zirconium alloys. An incident of pressure tube rupture due to the hydride blisters had happened in the Canadian CANDU reactor. This report describes the theoretical development and models on the formation and growth of hydride blister and some experimental results. The evaluation methodology and non-destructive testing for hydride blister in operating reactors are also described.

Extractive metallurgy of zirconium--1945 to the present

International Nuclear Information System (INIS)

Franklin, D.G.; Adamson, R.B.

1984-01-01

Although the history of the reduction of zirconium dates from 1824 and the first ductile zirconium metal was produced in the laboratory in 1914, modern reduction practice was pioneered by the U.S. Bureau of Mines starting in 1945. This paper reviews the history of the extractive metallurgy of zirconium from the early work of W. J. Kroll and co-workers at the Bureau of Mines in Albany, Ore., through the commercial development of the production of reactor-grade zirconium metal which was spurred by the requirements of the Naval Reactor Program and the development of commercial nuclear power. Technical subjects covered include processes for opening the ore, zirconium-hafnium separation, chlorination of zirconium oxide, reduction processes, and electrowinning of zirconium metal. Proposed new processes and process modifications are reviewed

In-reactor creep of zirconium alloys by thermal spikes

International Nuclear Information System (INIS)

Ibrahim, E.F.

1975-01-01

The size and duration of thermal spikes from fast neutrons have been calculated for zirconium alloys, showing that spikes up to 1.8 nm radius may exist for 2 x 10 -11 s at greater than melting point, at 570K ambient temperature. Creep rates have been calculated assuming that the elastic strain from the applied stress relaxes in the volume of the spikes (by preferential loop alignment or modification of an existing dislocation network). The calculated rates are consistent with strain rates observed in long term tests-in-reactor, if spike lifetimes are 2 to 2.5 x 10 -11 s. (Auth.)

A new model for the in-reactor corrosion of zirconium alloys

International Nuclear Information System (INIS)

Cox, B.

1997-01-01

Previous models for the in-reactor corrosion of zirconium alloys have assumed that the mechanism is a completely solid-state diffusion process, determined by the growth and breakdown of the protective oxide film. In-reactor kinetics have been related to out-reactor kinetics with the oxide-metal interface temperature calculated from effects of heat flux. Recent experimental results have suggested that oxide dissolution and reprecipitation may be a major process leading to the formation of thick porous oxide films in-reactor. The model described here is based on the dissolution of primary recoil tracks in the oxide as the primary process distinguishing in-reactor from out-reactor corrosion. The consequences of such a model would be a very different microscopic morphology of in-reactor and out-reactor thick films, a significant irradiation effect on non-heat transfer surfaces, and a change in the kinetics of the overall process. This model should be equally applicable to PWR and BWR water chemistries because of the amphoteric nature of ZrO 2 , and the effects of LiOH should operate by an essentially identical mechanism. A reciprocal rate equation should fit these processes and with additive terms seems capable of accommodating all water chemistry effects, except for discontinuous processes such as nodular corrosion. (author). 60 refs, 6 figs

Thermofluency in zirconium alloys

International Nuclear Information System (INIS)

Orozco M, E.A.

1976-01-01

A summary is presented about the theoretical and experimental results obtained at present in thermofluency under radiation in zirconium alloys. The phenomenon of thermofluency is presented in a general form, underlining the thermofluency at high temperature because this phenomenon is similar to the thermofluency under radiation, which ocurrs in zirconium alloys into the operating reactor. (author)

A new model for the in-reactor corrosion of zirconium alloys

Energy Technology Data Exchange (ETDEWEB)

Cox, B [University of Toronto, ON (Canada). Centre for Nuclear Engineering

1997-02-01

Previous models for the in-reactor corrosion of zirconium alloys have assumed that the mechanism is a completely solid-state diffusion process, determined by the growth and breakdown of the protective oxide film. In-reactor kinetics have been related to out-reactor kinetics with the oxide-metal interface temperature calculated from effects of heat flux. Recent experimental results have suggested that oxide dissolution and reprecipitation may be a major process leading to the formation of thick porous oxide films in-reactor. The model described here is based on the dissolution of primary recoil tracks in the oxide as the primary process distinguishing in-reactor from out-reactor corrosion. The consequences of such a model would be a very different microscopic morphology of in-reactor and out-reactor thick films, a significant irradiation effect on non-heat transfer surfaces, and a change in the kinetics of the overall process. This model should be equally applicable to PWR and BWR water chemistries because of the amphoteric nature of ZrO{sub 2}, and the effects of LiOH should operate by an essentially identical mechanism. A reciprocal rate equation should fit these processes and with additive terms seems capable of accommodating all water chemistry effects, except for discontinuous processes such as nodular corrosion. (author). 60 refs, 6 figs.

Challenges in design of zirconium alloy reactor components

International Nuclear Information System (INIS)

Kakodkar, Anil; Sinha, R.K.

1992-01-01

Zirconium alloy components used in core-internal assemblies of heavy water reactors have to be designed under constraints imposed by need to have minimum mass, limitations of fabrication, welding and joining techniques with this material, and unique mechanisms for degradation of the operating performance of these components. These constraints manifest as challenges for design and development when the size, shape and dimensions of the components and assemblies are unconventional or untried, or when one is aiming for maximization of service life of these components under severe operating conditions. A number of such challenges were successfully met during the development of core-internal components and assemblies of Dhruva reactor. Some of the then untried ideas which were developed and successfully implemented include use of electron beam welding, cold forming of hemispherical ends of reentrant cans, and a large variety of rolled joints of innovative designs. This experience provided the foundation for taking up and successfully completing several tasks relating to coolant channels, liquid poison channels and sparger channels for PHWRs and test sections for the in-pile loops of Dhruva reactor. For life prediction and safety assessment of coolant channels of PHWRs some analytical tools, notably, a computer code for prediction of creep limited life of coolant channels has been developed. Some of the future challenges include the development of easily replaceable coolant channels and also large diameter coolant channels for Advanced Heavy Water Reactor, and development of solutions to overcome deterioration of service life of coolant channels due to hydriding. (author). 5 refs., 13 figs., 1 tab

Waterside corrosion of zirconium alloys in nuclear power plants

International Nuclear Information System (INIS)

1998-01-01

Technically the study of corrosion of zirconium alloys in nuclear power reactors is a very active field and both experimental work and understanding of the mechanisms involved are going through rapid changes. As a result, the lifetime of any publication in this area is short. Because of this it has been decided to revise IAEA-TECDOC-684 - Corrosion of Zirconium Alloys in Nuclear Power Plants - published in 1993. This updated, revised and enlarged version includes major changes to incorporate some of the comments received about the first version. Since this review deals exclusively with the corrosion of zirconium and zirconium based alloys in water, and another separate publication is planned to deal with the fuel-side corrosion of zirconium based fuel cladding alloys, i.e. stress corrosion cracking, it was decided to change the original title to Waterside Corrosion of Zirconium Alloys in Nuclear Power Plants. The rapid changes in the field have again necessitated a cut-off date for incorporating new data. This edition incorporates data up to the end of 1995; including results presented at the 11 International Symposium on Zirconium in the Nuclear Industry held in Garmisch-Partenkirchen, Germany, in September 1995. The revised format of the review now includes: Introductory chapters on basic zirconium metallurgy and oxidation theory; A revised chapter discussing the present extent of our knowledge of the corrosion mechanism based on laboratory experiments; a separate and revised chapter discussing hydrogen uptake; a completely reorganized chapter summarizing the phenomenological observations of zirconium alloy corrosion in reactors; a new chapter on modelling in-reactor corrosion; a revised chapter devoted exclusively to the manner in which irradiation might influence the corrosion process; finally, a summary of our present understanding of the corrosion mechanisms operating in reactor

Atomic hydrogen reactor

International Nuclear Information System (INIS)

Massip de Turville, C.M.D.

1982-01-01

Methods are discussed of generating heat in an atomic hydrogen reactor which involve; the production of atomic hydrogen by an electrical discharge, the capture of nascent neutrons from atomic hydrogen in a number of surrounding steel alloy tubes having a high manganese content to produce 56 Mn, the irradiation of atomic hydrogen by the high energy antineutrinos from the beta decay of 56 Mn to yield nascent neutrons, and the removal of the heat generated by the capture of nascent neutrons by 55 Mn and the beta decay of 56 Mn. (U.K.)

Anisotropy of mechanical properties of zirconium and zirconium alloys

International Nuclear Information System (INIS)

Medrano, R.E.

1975-01-01

In studies of technological applications of zirconium to fuel elements of nuclear reactor, it was found that the use of plasticity equations for isotropic materials is not in agreement with experimental results, because of the strong anisotropy of zirconium. The present review describes recent progress on the knowledge of the influence of anisotropy on mechanical properties, after Douglass' review in 1971. The review was written to be selfconsistent, changing drastically the presentation of some of the referenced papers. It is also suggested some particular experiments to improve developments in this area

Automatic measuring system of zirconium thickness for zirconium liner cladding tubes

International Nuclear Information System (INIS)

Matsui, K.; Yamaguchi, H.; Hiroshima, T.; Sakamoto, T.; Murayama, R.

1985-01-01

An automatic system of pure zirconium liner thickness for zirconium-zircaloy cladding tubes has been successfully developed. The system consists of three parts. (1) An ultrasonic thickness measuring method for mother tubes before cold rolling. (2) An electromagnetic thickness measuring method for the manufactured tubes. (3) An image processing method for the cross sectional view of the manufactured cut tube samples. In Japanese nuclear industry, zirconium-zircaloy cladding tubes have been tested in order to realize load following operation in the atomic power plant. In order to provide for the practical use in the near future, Sumitomo Metal Industries, Ltd. has been studied and established the practical manufacturing process of the zirconium liner cladding tubes. The zirconium-liner cladding tube is a duplex tube comprising an inner layer of pure zirconium bonded to zircaloy metallurgically. The thickness of the pure zirconium is about 10 % of the total wall thickness. Several types of the automatic thickness measuring methods have been investigated instead of the usual microscopic viewing method in which the liner thickness is measured by the microscopic cross sectional view of the cut tube samples

Development of zirconium hydride highly effective moderator materials

International Nuclear Information System (INIS)

Yin Changgeng

2005-10-01

The zirconium hydride with highly content of hydrogen and low density is new efficient moderator material for space nuclear power reactor. Russia has researched it to use as new highly moderator and radiation protection materials. Japanese has located it between the top of pressure vessel and the main protection as a shelter, the work temperature is rach to 220 degree C. The zirconium hydride moderator blocks are main parts of space nuclear power reactor. Development of zirconium hydride moderator materials have strength research and apply value. Nuclear Power Research and Design Instituteoh China (NPIC) has sep up the hydrogenation device and inspect systems, and accumurate a large of experience about zirconium hydride, also set up a strict system of QA and QC. (authors)

Analysis of hafnium in zirconium alloys

International Nuclear Information System (INIS)

Kondo, Isao; Sakai, Fumiaki; Ohuchi, Yoshifusa; Nakamura, Hisashi

1977-01-01

It is required to analyse alloying components and impurity elements in the acceptance analysis of zirconium alloys as the material for fuel cladding tubes and pressure tubes for advanced thermal reactors. Because of extreme similarity in chemical properties between zirconium and hafnium, about 100 ppm of hafnium is usually contained in zirconium alloys. Zircaloy-2 alloy and 2.5% Nb-zirconium with the addition of hafnium had been prepared as in-house standard samples for rapid analysis. Study was made on fluorescent X-ray analysis and emission spectral analysis to establish the analytical method. By using these in-house standard samples, acceptance analysis was successfully carried out for the fuel cladding tubes for advanced thermal reactors. Sulfuric acid solution was prepared from JAERI-Z 1, 2 and 3, the standard sample for zircaloy-2 prepared by the Analytical Committee on Nuclear Fuel and Reactor Materials, JAERI, and zirconium oxide (Hf 1 ppm/Zr). Standard Hf solution was added to the sulfuric acid solution step by step, to make up a series of the standard oxide samples by the precipitation process. By the use of these standard samples, the development of the analytical method and joint analysis were made by the three-member analytical technique research group including PNC. The analytical precision for the fluorescent X-ray analysis was improved by attaching a metallic yttrium filter to the window of an X-ray tube so as to suppress the effect due to zirconium matrix. The variation factor of the joint analysis was about 10% to show good agreement, and the indication value was determined. (Kobatake, H.)

Lithium aluminate/zirconium material useful in the production of tritium

Science.gov (United States)

Cawley, W.E.; Trapp, T.J.

A composition is described useful in the production of tritium in a nuclear reactor. Lithium aluminate particles are dispersed in a matrix of zirconium. Tritium produced by the reactor of neutrons with the lithium are absorbed by the zirconium, thereby decreasing gas pressure within capsules carrying the material.

Zirconium - ab initio modelling of point defects diffusion

International Nuclear Information System (INIS)

Gasca, Petrica

2010-01-01

Zirconium is the main element of the cladding found in pressurized water reactors, under an alloy form. Under irradiation, the cladding elongate significantly, phenomena attributed to the vacancy dislocation loops growth in the basal planes of the hexagonal compact structure. The understanding of the atomic scale mechanisms originating this process motivated this work. Using the ab initio atomic modeling technique we studied the structure and mobility of point defects in Zirconium. This led us to find four interstitial point defects with formation energies in an interval of 0.11 eV. The migration paths study allowed the discovery of activation energies, used as entry parameters for a kinetic Monte Carlo code. This code was developed for calculating the diffusion coefficient of the interstitial point defect. Our results suggest a migration parallel to the basal plane twice as fast as one parallel to the c direction, with an activation energy of 0.08 eV, independent of the direction. The vacancy diffusion coefficient, estimated with a two-jump model, is also anisotropic, with a faster process in the basal planes than perpendicular to them. Hydrogen influence on the vacancy dislocation loops nucleation was also studied, due to recent experimental observations of cladding growth acceleration in the presence of this element [fr

Zirconium alloy barrier having improved corrosion resistance

International Nuclear Information System (INIS)

Adamson, R.B.; Rosenbaum, H.S.

1983-01-01

A nuclear fuel element for use in the core of a nuclear reactor has a composite cladding container having a substrate and a dilute zirconium alloy liner bonded to the inside surface of the substrate. The dilute zirconium alloy liner forms about 1 to about 20 percent of the thickness of the cladding and is comprised of zirconium and a metal selected from the group consisting of iron, chromium, iron plus chromium, and copper. The dilute zirconium alloy liner shields the substrate from impurities or fission products from the nuclear fuel material and protects the substrate from stress corrosion and stress cracking. The dilute zirconium alloy liner displays greater corrosion resistance, especially to oxidation by hot water or steam than unalloyed zirconium. The substrate material is selected from conventional cladding materials, and preferably is a zirconium alloy. (author)

Verification of codes used for the nuclear safety assessment of the small space heterogeneous reactors with zirconium hydride moderator

International Nuclear Information System (INIS)

Glushkov, E.S.; Gomin, E.A.; Kompaniets, G.V.

1994-01-01

Computer codes used for assessment of nuclear safety for space NPP are compared taking as an example small-sized heterogeneous reactor with zirconium hydride moderator of the Topaz-2 facility. The code verifications are made for five different variants

Electron microscopy of nuclear zirconium alloys

International Nuclear Information System (INIS)

Versaci, R.A.; Ipohorski, Miguel

1986-01-01

Transmission electron microscopy observations of the microstructure of zirconium alloys used in fuel sheaths of nuclear power reactors are reported. Specimens were observed after different thermal and mechanical treatment, similar to those actually used during fabrication of the sheaths. Electron micrographs and electron diffraction patterns of second phase particles present in zircaloy-2 and zircaloy-4 were also obtained, as well as some characteristic parameters. Images of oxides and hydrides most commonly present in zirconium alloys are also shown. Finally, the structure of a Zr-2,5Nb alloy used in CANDU reactors pressure tubes, is observed by electron microscopy. (Author) [es

Influence of preliminary radiation-oxidizing treatment on the corrosion resistance of zirconium in conditions of action of ionizing radiation

International Nuclear Information System (INIS)

Garibov, A. A.; Aliyev, A. G.; Agayev, T. N.; Velibekova, G. Z.

2004-01-01

Today mainly water-cooled nuclear reactors predominate in atomic energetics. For safe work of nuclear reactors detection of accumulation process of explosives, formed during radiation and temperature influence on heat-carriers in contact with materials of nuclear reactors in normal and emergency regimes of work is of great importance. The main sources of molecular hydrogen formation in normal and emergency regimes are the processes of liquid and vaporous water in vapo metallic reaction [1-5]. At the result of these processes molecular hydrogen concentration in heat-carrier composition always exceeds theoretically expected concentration. One of the main ways to solve the problem of water-cooled reactors safety is detection of possibilities to raise material resistance of fuel elements and heat carrier to joint action of ionizing radiation and temperature. The second way is inhibition of radiation-catalytic activity of construction materials' surface during the process of water decomposition. It's been established, that one of the ways to raise resistance of zirconium materials to the influence of ionizing radiation is formation of thin oxide film on the surface of metals. In the given work the influence of preliminary oxidizing treatment of zirconium surface on its radiation-catalytic activity during the process of water decomposition. With this aim zirconium is exposed to preliminary influence of gamma-quantum in contact with hydrogen peroxide at different meanings of absorbed radiation dose

Radiochemical neutron activation analysis of zirconium and zirconium-niobium alloys

International Nuclear Information System (INIS)

Tashimova, F.A.; Sadikov, I.I.; Salimov, M.

2004-01-01

Full text: Zirconium and zirconium-niobium alloys are used on nuclear technology, as fuel cladding of nuclear reactors. Their nuclear-physical, mechanical and thermophysical properties are influenced them matrix and impurity composition, therefore determination of matrix and impurity content of these materials is a very important task. Neutron activation analysis is one from multielemental and high sensible techniques that are widely applied in analysis of high purity materials. Investigation of nuclear-physical characteristics of zirconium has shown that instrumental variant NAA is unusable for analysis due to high radioactivity of a matrix. Therefore it is necessary carrying out radiochemical separation of impurity radionuclides from matrix. Study of the literature datum have shown, that zirconium and niobium are very well extracted from muriatic solution with 5% tributyl phosphineoxide (TBPO) solution in toluene and 0,75 M solution of di-2-ethyl hexyl phosphoric acid (HDEHP) in cyclohexanone. Investigation of these elements extraction in these systems has shown that more effective and selective separation of matrix radionuclides is achieved in HDEHP-3M HCI system. This system is also extracted and hafnium, witch is an accompanying element of zirconium and its high content prevented determination of other impurity elements in sample. Therefore we used extraction system HDEHP-3M HCl for analysis of zirconium and zirconium-niobium alloys in chromatographic variant. By measurement of distribution profile of a matrix and of elution curve of determined elements is established, that for effective separation of impurity and matrix radionuclides there is enough chromatographic column with diameter 1 cm and height of a sorbent layer 7 cm, thus volume of elute, necessary for complete elution of determinate elements is 35-40 ml. On the basis of the carried out researches the technique of radiochemical NAA of high purity zirconium and zirconium-niobium alloy, which allows to

Quantum chemical study of the elementary reactions in zirconium oxide atomic layer deposition

International Nuclear Information System (INIS)

Widjaja, Yuniarto; Musgrave, Charles B.

2002-01-01

Elementary reactions in atomic layer deposition of zirconia using zirconium tetrachloride and water are investigated using the density functional theory. The atomistic mechanisms of the two deposition half cycles on the Zr-OH and Zr-Cl surface sites are investigated. Both half reactions proceed through the formation of stable intermediates, resulting in high barriers for HCl formation. We find that the intermediate stability is lowered as the surface temperature is raised. However, increasing temperature also increases the dissociation free-energy barrier, which in turn results in increased desorption of adsorbed precursors

Irradiation effects on Zr-2.5Nb in power reactors

Energy Technology Data Exchange (ETDEWEB)

Song, C., E-mail: Carol.Song@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

2016-06-15

Zirconium alloys are widely used as structural materials in nuclear applications because of their attractive properties such as a low absorption cross-section for thermal neutrons, excellent corrosion resistance in water, and good mechanical properties at reactor operating temperatures. Zr-2.5Nb is one of the most commonly used zirconium alloys and has been used for pressure tube materials in CANDU (Canada Deuterium Uranium) and RBMK (Reaktor Bolshoy Moshchnosti Kanalnyy, 'High Power Channel-type Reactor') reactors for over 40 years. In a recent report from the Electric Power Research Institute, Zr-2.5Nb was identified as one of the candidate materials for use in normal structural applications in light-water reactors owing to its increased resistance to irradiation-induced degradation as compared with currently used materials. Historically, the largest program of in-reactor tests on zirconium alloys was performed by Atomic Energy of Canada Limited. Over many years of in-reactor testing and CANDU operating experience with Zr- 2.5Nb, extensive research has been conducted on the irradiation effects on its microstructures, mechanical properties, deformation behaviours, fracture toughness, delayed hydride cracking, and corrosion. Most of the results on Zr-2.5Nb obtained from CANDU experience could be used to predict the material performance under light water reactors. This paper reviews the irradiation effects on Zr-2.5Nb in power reactors (including heavy-water and light-water reactors) and summarizes the current state of knowledge. (author)

Fabrication of high performance components for Indian nuclear reactors

International Nuclear Information System (INIS)

Jayaraj, R.N.

2011-01-01

Nuclear Fuel Complex (NFC), a Unit of the Department of Atomic Energy (DAE) has been engaged for well over three-and-half decades in the manufacture of fuels for Pressurized Heavy Water Reactors (PHWRs) and Boiling Water Reactors (BWRs). All the fuel assembly components, like, fuel clad tubes, end plugs, spacers, spacer grids etc. are also being manufactured at NFC in Zirconium alloy material. Apart from the regular production of these components and finished fuel assemblies, NFC has also been engaged in the production of Zirconium alloy reactor core structurals, like, pressure tubes, calandria tubes, garter springs and reactivity control mechanisms for PHWRs and square channels for BWRs. While all these structural components are produced through standardized flow sheets, there have been continuous innovations carried out in the processes to meet the ever increasing end-use characteristics laid down by the utilities. The paper enumerates various aspects of different technologies developed at NFC for the manufacture of high performance components for reactor applications

Zirconium - an imported mineral commodity

International Nuclear Information System (INIS)

1983-10-01

This report examines Canada's position in regard to the principal zirconium materials: zircon; fusion-cast zirconium-bearing refractory products; zirconium-bearing chemicals; and zirconium metal, master alloys, and alloys. None of these is produced in Canada except fused alumina-zirconia and certain magnesium-zirconium alloys and zirconium-bearing steels. Most of the 3 000-4 000 tonnes of the various forms of zircon believed to be consumed in Canada each year is for foundry applications. Other minerals, notably chromite, olivine and silica sand are also used for these purposes and, if necessary, could be substituted for zircon. Zirconium's key role in Canada is in CANDU nuclear power reactors, where zirconium alloys are essential in the cladding for fuel bundles and in capital equipment such as pressure tubes, calandria tubes and reactivity control mechanisms. If zirconium alloys were to become unavailable, the Canadian nuclear power industry would collapse. As a contingency measure, Ontario Hydro maintains at least nine months' stocks of nuclear fuel bundles. Canada's vulnerability to short-term disruptions to supplies of nuclear fuel is diminished further by the availability of more expensive electricity from non-nuclear sources and, given time, from mothballed thermal plants. Zirconium minerals are present in many countries, notably Australia, the Republic of South Africa and the United States. Australia is Canada's principal source of zircon imports; South Africa is its sole source of baddeleyite. At this time, there are no shortages of either material. Canada has untapped zirconium resources in the Athabasca Oil Sands (zircon) and at Strange Lake along the ill-defined border between Quebec and Newfoundland (gittinsite). Adequate metal and alloy production facilities exist in France, Japan and the United States. No action by the federal government in regard to zirconium supplies is called for at this time

New solvent extraction process for zirconium and hafnium

International Nuclear Information System (INIS)

Takahashi, M.; Katoh, Y.; Miyazaki, H.

1984-01-01

The authors' company developed a new solvent extraction process for zirconium and hafnium separation, and started production of zirconium sponge by this new process in September 1979. The process utilizes selective extraction of zirconium oxysulfate using high-molecular alkyl amine, and has the following advantages: 1. This extraction system has a separation factor as high as 10 to 20 for zirconium and hafnium in the range of suitable acid concentration. 2. In the scrubbing section, removal of all the hafnium that coexists with zirconium in the organic solvent can be effectively accomplished by using scrubbing solution containing hafnium-free zirconium sulfate. Consequently, hafnium in the zirconium sponge obtained is reduced to less than 50 ppm. 3. The extractant undergoes no chemical changes but is very stable for a long period. In particular, its solubility in water is small, about 20 ppm maximum, posing no environmental pollution problems such as are often caused by other process raffinates. At the present time, the zirconium and hafnium separation operation is very stable, and zirconium sponge made by this process can be applied satisfactorily to nuclear reactors

Determination of hydrogen in zirconium hydride and uranium-zirconium hydride by inert gas exraction-gravimetric method

International Nuclear Information System (INIS)

Hoshino, Akira; Iso, Shuichi

1976-01-01

An inert gas extraction-gravimetric method has been applied to the determination of hydrogen in zirconium hydride and uranium-zirconium hydride which are used as neutron moderator and fuel of nuclear safety research reactor (NSRR), respectively. The sample in a graphite-enclosed quartz crucible is heated inductively to 1200 0 C for 20 min in a helium stream. Hydrogen liberated from the sample is oxidized to water by copper(I) oxide-copper(II) oxide at 400 0 C, and the water is determined gravimetrically by absorption in anhydrone. The extraction curves of hydrogen for zirconium hydride and uranium-zirconium hydride samples are shown in Figs. 2 and 3. Hydrogen in the samples is extracted quantitatively by heating at (1000 -- 1250) 0 C for (10 -- 40) min. Recoveries of hydrogen in the case of zirconium hydride were examined as follows: a weighed zirconium rod (5 phi x 6 mm, hydrogen -5 Torr. After the chamber was filled with purified hydrogen to 200 Torr, the rod was heated to 400 0 C for 15 h, and again weighed to determine the increase in weight. Hydrogen in the rod was then determined by the proposed method. The results are in excellent agreement with the increase in weight as shown in Table 1. Analytical results of hydrogen in zirconium hydride samples and an uranium-zirconium hydride sample are shown in Table 2. (auth.)

Synthesis of Zirconium Lower Chlorides

International Nuclear Information System (INIS)

Gaviria, Juan P.

2002-01-01

This research is accurately related to the Halox concept of research reactor spent fuel element treatment.The aim of this project is to work the conditioning through selected chlorination of the element that make the spent fuel element. This research studied the physical chemistry conditions which produce formation of the lower zirconium chlorides through the reaction between metallic Zr and gaseous ZrCl 4 in a silica reactor.This work focused special attention in the analysis and confrontation of the published results among the different authors in order to reveal coincidences and contradictions.Experimental section consisted in a set of synthesis with different reaction conditions and reactor design. After reaction were analyzed the products on Zr shavings and the deposit growth on wall reactor.The products were strongly dependent of reactor design. It was observed that as the distance between Zr and wall reactor increased greater was tendency to lower chlorides formation.In reactors with small distance the reaction follows other way without formation of lower chlorides.Analysis on deposit growth on reactor showed that may be formed to a mixture of Si x Zr y intermetallics and zirconium oxides.Presence of oxygen in Zr and Zr-Si compounds on wall reactor reveals that there is an interaction between quartz and reactants.This interaction is in gaseous phase because contamination is observed in experiences where Zr was not in contact with reactor.Finally, it was made a global analysis of all experiences and a possible mechanism that interprets reaction ways is proposed

Use of research reactors in Soviet atomic centres

International Nuclear Information System (INIS)

1964-01-01

The manner of controlling and directing research reactors in the USSR was described in October at the IAEA seminar for atomic energy administrators by Dr. U. V. Archangelski, Department Chief, State Committee for Utilization of Atomic Energy, USSR. He also enumerated the research reactors in operation. In addition to the portions of the paper which are quoted below, he gave details of the scientific work being carried out in these reactors.

Production of nuclear grade zirconium: A review

Energy Technology Data Exchange (ETDEWEB)

Xu, L., E-mail: L.Xu-2@tudelft.nl [School of Materials Science and Metallurgy, Northeastern University, Shenyang 110004 (China); Department of Materials Science and Engineering, Delft University of Technology, Delft 2628 CD (Netherlands); Xiao, Y. [Department of Metallurgical Engineering, Anhui University of Technology, Ma' anshan 243002 (China); Zr-Hf-Ti Metallurgie B.V., Den Haag 2582 SB (Netherlands); Sandwijk, A. van [Zr-Hf-Ti Metallurgie B.V., Den Haag 2582 SB (Netherlands); Xu, Q. [School of Materials Science and Metallurgy, Northeastern University, Shenyang 110004 (China); Yang, Y. [Department of Materials Science and Engineering, Delft University of Technology, Delft 2628 CD (Netherlands)

2015-11-15

Zirconium is an ideal material for nuclear reactors due to its low absorption cross-section for thermal neutrons, whereas the typically contained hafnium with strong neutron-absorption is very harmful for zirconium as a fuel cladding material. This paper provides an overview of the processes for nuclear grade zirconium production with emphasis on the methods of Zr–Hf separation. The separation processes are roughly classified into hydro- and pyrometallurgical routes. The known pyrometallurgical Zr–Hf separation methods are discussed based on the following reaction features: redox characteristics, volatility, electrochemical properties and molten salt–metal equilibrium. In the present paper, the available Zr–Hf separation technologies are compared. The advantages and disadvantages as well as future directions of research and development for nuclear grade zirconium production are discussed.

Radiation induced defect flux behaviors at zirconium based component

International Nuclear Information System (INIS)

Choi, Sang Il; Kim, Ji Hyun; Kwon, Jun Hyun; Lee, Gyeong Geun

2013-01-01

In commercial reactor core, structure materials are located in high temperature and high pressure environment. Therefore, main concern of structure materials is corrosion and mechanical properties change than radiation effects on materials. However, radiation effects on materials become more important phenomena because research reactor condition is different from commercial reactor. The temperature is lower than 100 .deg. C and radiation dose is much higher than that of commercial reactor. Among the radiation effect on zirconium based metal, radiation induced growth (RIG), known as volume conservative distortion, is one of the most important phenomena. Recently, theoretical RIG modeling based on radiation damage theory (RDT) and balance equation are developed. However, these growth modeling have limited framework of single crystal and high temperature. To model theoretical RIG in research reactor, qualitative mechanism must be set up. Therefore, this paper intent is establishing defect flux mechanism of zirconium base metal in research reactor for RIG modeling. After than theoretical RIG work will be expanded to research reactor condition

Experimental and thermodynamic study of the erbium-oxygen-zirconium and gadolinium-oxygen-zirconium systems

International Nuclear Information System (INIS)

Jourdan, J.

2009-11-01

This work is a contribution to the development of innovative concepts for fuel cladding in pressurized water nuclear reactors. This concept implies the insertion of rare earth (erbium and gadolinium) in the zirconium fuel cladding. The determination of phase equilibria in the systems is essential prior to the implementation of such a promising solution. This study consisted in an experimental determination of the erbium-zirconium phase diagram. For this, we used many different techniques in order to obtain diagram data such as solubility limits, solidus, liquidus or invariant temperatures. These data allowed us to present a new diagram, very different from the previous one available in the literature. We also assessed the diagram using the CALPHAD approach. In the gadolinium-zirconium system, we determined experimentally the solubility limits. Those limits had never been determined before, and the values we obtained showed a very good agreement with the experimental and assessed versions of the diagram. Because these alloys are subjected to oxygen diffusion throughout their life, we focused our attention on the erbium-oxygen-zirconium and gadolinium-oxygen-zirconium systems. The first system has been investigated experimentally. The alloys fabrication has been performed using powder metallurgy. In order to obtain pure raw materials, we fabricated powder from erbium and zirconium bulk metals using hydrogen absorption/desorption. The characterisation of the ternary pellets allowed the determination of two ternary isothermal sections at 800 and 1100 C. For the gadolinium-oxygen-zirconium system, we calculated the phase equilibria at temperatures ranging from 800 to 1100 C, using a homemade database compiled from literature assessments of the oxygen-zirconium, gadolinium-zirconium and gadolinia-zirconia systems. Finally, we determined the mechanical properties, in connexion with the microstructure, of industrial quality alloys in order to identify the influence of

In-reactor creep of zirconium-2.5 wt% niobium at 570 K

International Nuclear Information System (INIS)

Coleman, C.E.; Causey, A.R.; Fidleris, V.

1976-01-01

The effect of fast neutron flux at 570 K on the creep rate of specimens of zirconium-2.5 wt% niobium alloy taken from tubes in various metallurgical conditions has been measured using both constant load tensile creep machines and bent-beam stress relaxation. Creep rates calculated from stress relaxation fit on the trend line for the constant load creep data. Between 114 MPa and 450 MPa the creep rate is proportional to neutron flux. The creep rate of specimens from the longitudinal direction is about twice that of specimens from the circumferential direction of a tube. This anisotropy in creep strength is attributed partly to crystallographic texture and partly to deformation substructure. Cold-work is detrimental to in-reactor creep strength; as-extruded material has higher creep strength. In cold-worked material at stresses below 100 MPa the stress exponent, n, is about 1; n gradually increases with stress being about 10 at 525 MPa and about 100 at 660 MPa. In laboratory tests, rupture ductility correlates inversely with n; the lower n the higher the ductility. In-reactor tests support this correlation thus pressure tubes in CANDU reactors, operating at 117 MPa where n approximately 1, should have good ductility. (Auth.)

DISSOLUTION OF ZIRCONIUM AND ALLOYS THEREFOR

Science.gov (United States)

Swanson, J.L.

1961-07-11

The dissolution of zirconium cladding in a water solution of ammonium fluoride and ammonium nitrate is described. The method finds particular utility in processing spent fuel elements for nuclear reactors. The zirconium cladding is first dissolved in a water solution of ammonium fluoride and ammonium nitrate; insoluble uranium and plutonium fiuorides formed by attack of the solvent on the fuel materiai of the fuel element are then separated from the solution, and the fuel materiai is dissolved in another solution.

Determination of the population of octahedral and tetrahedral interstitials in zirconium hydrides

International Nuclear Information System (INIS)

Fedorov, V.M.; Gogava, V.V.; Shilo, S.I.; Biryukova, E.A.

1983-01-01

Results of neutron investigations of ZrHsub(1.66), ZrHsub(1.75) and ZrHsub(1.98) zirconium hydrides are presented. Investigations were conducted using plane polycrystal samples by multidetector system of scattered neutron detection. Neutron diffraction method was used to determine the number of interstitial hydrogen atoms in interstitials of the lattice cell in the case of statistic atom distribution. The numbers of interstitial atoms in octahedral interstitials for zirconium hydrides were determined experimentally; the difference of potential energies of hydrogen atoms in octa- and tetrahedral interstitials was determined as well. It is shown that experimentally determined difference of potential energies of hydrogen atoms, occupying octa- and tetrahedral positions in investigated zirconium hydrides results at room temperature in the pretailing occupation of tetrahedral interstitials by hydrogen atoms (85-90%); the occupation number grows with temperature decrease and the ordering of interstitial vacancies with formation of hydrogen superstructure takes place at low temperatures

Atomistic modeling of zirconium hydride precipitation: methodology for deriving a tight-binding potential

International Nuclear Information System (INIS)

Dufresne, Alice

2014-01-01

The zirconium-hydrogen system is of nuclear safety interest, as the hydride precipitation leads to the cladding embrittlement, which is made of zirconium-based alloys. The cladding is the first safety barrier confining the radioactive products: its integrity shall be kept during the entire fuel-assemblies life, in reactor, including accidental situation, and post-operation (transport and storage). Many uncertainties remain regarding the hydrides precipitation kinetics and the local stress impact on their precipitation. The atomic scale modeling of this system would bring clarifications on the relevant mechanisms. The usual atomistic modeling methods are based on thermo-statistic approaches, whose precision and reliability depend on the interatomic potential used. However, there was no potential allowing a rigorous study of the Zr-H system. The present work has indeed addressed this issue: a new tight-binding potential for zirconium hydrides modeling is now available. Moreover, this thesis provides a detailed manual for deriving such potentials accounting for spd hybridization, and fitted here on DFT results. This guidebook has be written in light of modeling a pure transition metal followed by a metal-covalent coupling (metallic carbides, nitrides and silicides). (author)

Titanium zirconium and hafnium coordination compounds with vanillin thiosemicarbazone

International Nuclear Information System (INIS)

Konunova, Ts.B.; Kudritskaya, S.A.

1987-01-01

Coordination compounds of titanium zirconium and hafnium tetrachlorides with vanillin thiosemicarbazone of MCl 4 x nLig composition, where n=1.5, 4 for titanium and 1, 2, 4 for zirconium and hafnium, are synthesized. Molar conductivity of ethanol solutions is measured; IR spectroscopic and thermochemical investigation are carried out. The supposition about ligand coordination via sulfur and azomethine nitrogen atoms is made. In all cases hafnium forms stable compounds than zirconium

Determination of zirconium by fluoride ion selective electrode

International Nuclear Information System (INIS)

Mahanty, B.N.; Sonar, V.R.; Gaikwad, R.; Raul, S.; Das, D.K.; Prakash, A.; Afzal, Md.; Panakkal, J.P.

2010-01-01

Full text: Zirconium is used in a wide range of applications including nuclear clad, catalytic converters, surgical appliances, metallurgical furnaces, superconductors, ceramics, lamp filaments, anti corrosive alloys and photographical purposes. Irradiation testing of U-Zr and U-Pu-Zr fuel pins has also demonstrated their feasibility as fuel in liquid metal reactors. Different methods that are employed for the determination of zirconium are spectrophotometry, potentiometry, neutron activation analysis and mass spectrometry. Ion-selective electrode (ISE), selective to zirconium ion has been studied for the direct potentiometric measurements of zirconium ions in various samples. In the present work, an indirect method has been employed for the determination of zirconium in zirconium nitrate sample using fluoride ion selective electrode. This method is based on the addition of known excess amount of fluoride ion to react with the zirconium ion to produce zirconium tetra fluoride at about pH 2-3, followed by determination of residual fluoride ion selective electrode. The residual fluoride ion concentrations were determined from the electrode potential data using calibration plot. Subsequently, zirconium ion concentrations were determined from the concentration of consumed fluoride ions. A precision of about 2% (RSD) with the mean recovery of more than 94% has been achieved for the determination of zirconium at the concentration of 4.40 X 10 -3 moles lit -1

Zirconium alloy fuel cladding resistant to PCI crack propagation

International Nuclear Information System (INIS)

Boyle, R.F.; Foster, J.P.

1987-01-01

A nuclear fuel element is described cladding tube comprising: concentric tubular layers of zirconium base alloys; the concentric tubular layers including an inner layer and outer layer; the outer layer metallurgically bonded to the inner layer; the outer layer composed of a first zirconium base alloy characterized by excellent resistance to corrosion caused by exposure to high temperature and pressure aqueous environments; the inner layer composed of a second zirconium base alloy consisting of: about 0.2 to 0.6 wt.% tin, about 0.03 to 0.11 wt.% iron, less than about 0.02 wt.% chromium, up to about 350 ppm oxygen and the remainder being zirconium and incidental impurities, and the inner layer characterized by improved resistance to crack propagation under reactor operating conditions compared to the first zirconium alloy

Titanium(IV), zirconium, hafnium and thorium

International Nuclear Information System (INIS)

Brown, Paul L.; Ekberg, Christian

2016-01-01

Titanium can exist in solution in a number of oxidation states. The titanium(IV) exists in acidic solutions as the oxo-cation, TiO 2+ , rather than Ti 4+ . Zirconium is used in the ceramics industry and in nuclear industry as a cladding material in reactors where its reactivity towards hydrolysis reactions and precipitation of oxides may result in degradation of the cladding. In nature, hafnium is found together with zirconium and as a consequence of the contraction in ionic radii that occurs due to the 4f -electron shell, the ionic radius of hafnium is almost identical to that of zirconium. All isotopes of thorium are radioactive and, as a consequence of it being fertile, thorium is important in the nuclear fuel cycle. The polymeric hydrolysis species that have been reported for thorium are somewhat different to those identified for zirconium and hafnium, although thorium does form the Th 4 (OH) 8 8+ species.

Some recent trends in the use of zirconium alloys for nuclear service

International Nuclear Information System (INIS)

Balaramamoorthy, K.

1992-01-01

Without any exception nuclear power reactors particularly the water cooled ones, operating in the World use natural or slightly enriched uranium oxide fuel pellets with zirconium alloy cladding. While the zirconium alloys have proven to be successful in their designed usage, a desire for longer lifetimes of core components and increased duty cycle puts more demand on materials performance. This demand has led to more in depth studies of phenomena associated with zirconium alloy corrosion mechanism, fine tuning of the zirconium alloy composition, development of fabrication techniques and to the evaluation of newer zirconium alloys for critical applications. (author). 5 refs., 32 figs

Process for surface treatment of zirconium-containing cladding materials for fuel element or other components for nuclear reactors

International Nuclear Information System (INIS)

Videm, K.G.; Lunde, L.R.; Kooyman, H.H.

1975-01-01

A process for the surface treatment of zirconium-base cladding materials for fuel elements or other components for nuclear reactors is described. The treatment includes pickling the cladding material in a fluoride-containing bath, and then applying a protective coating through oxidation to the pickled cladding material. The fluoride-containing contaminants which remain on the surface of the cladding material during pickling are removed or rendered harmless by anodic oxidation

In-pile creep test technique for zirconium alloys examination in BR-10 reactor channels

International Nuclear Information System (INIS)

Pevchikh, Yu.M.; Kruglov, A.S.; Troyanov, V.M.

2002-01-01

The irradiation enhanced creep phenomenon was discovered in stainless steels as a specific physical process accompanying high-intensity neutron flux irradiation in fast reactors. IPPE is also experienced in irradiation creep test activities, studying different types of materials under irradiation in BR-10 fast reactor. Series of in-channel type test facilities were constructed and tested in BR-10 reactor's 'dry' channels in order to carry out full-scale instrumented examination regarded to in-pile creep behaviour of different reactor materials. As a result, a specific test technique, named 'Tensometric method', has been developed and experimentally proved to be power enough in order to investigate irradiation creep of materials right in situ under neutron irradiation. The main peculiarity of test facility, which is constructed to apply the tensometric method, consists in absence of any special deformation-measurement cell at all. The in-pile creep strain measurement technique developed at IPPE is based on the non-direct measurement of specimen's deformation (either linear tensile strain or angular twisting one), which directly affects the loaded draws' tension parameters. Starting from 1993, in-pile creep experiments to investigate in-reactor creep behaviour of E110 and E635 zirconium alloys were carried out in BR-10. Experimental results and data collected during more than 20-year of BR-10 in-reactor creep test experience can be assumed as a strong evidence that the tensometric technique is a powerful instrument, which can give a chance to study different irradiation effects on reactor materials directly under irradiation. (author)

ICP-AES determination of rare earths in zirconium with prior chemical separation of the matrix

International Nuclear Information System (INIS)

Rajeswari, B.; Dhawale, B.A.; Page, A.G.; Sastry, M.D.

2002-01-01

Zirconium being one of the most important material in nuclear industry used as a fuel cladding in reactors and an additive in advanced fuels necessitates its characterization for trace metallic contents. Zirconium, as refractory in nature as the rare earth elements, has a complex spectrum comprising of several emission lines. Rare earths, which are high neutron absorbers have to be analysed at very low limits. Hence, to achieve the desired limits, the major matrix has to be separated prior to rare earth determination. The present paper describes the method developed for the separation of rare earths from zirconium by solvent extraction using Trioctyl Phosphine Oxide (TOPO) as the extractant followed by their determination using Inductively Coupled Plasma - Atomic Emission Spectrometric (ICP-AES) method. Initially, radiochemical studies were carried out using known amounts of gamma active tracers of 141 Ce, 152-154 Eu, 153 Gd and 95 Zr for optimisation of extraction conditions using Tl- activated NaI detector. The optimum conditions at 0.5 M TOPO/xylene in 6 M HCl so as to achieve a quantitative recovery of rare earth analytes alongwith a near total extraction of zirconium in the organic phase, was further extended to carry out the studies using ICP-AES method. The recovery of rare earths was found to be quantitative within experimental error with a precision better than 10% RSD. (author)

A half-century of changes in zirconium alloys

International Nuclear Information System (INIS)

Mardon, J.P.; Barberis, P.; Hoffmann, P.B.

2008-01-01

This article presents the history of zirconium alloys for PWR and BWR technologies. For more than 20 years zirconium alloys have evolved to cope with demands of the reactor operators concerning the burn-up extension and new safety margins. The poor properties of Zircaloy-1 concerning corrosion have led researchers to add elements like iron by developing Zircaloy-3A and Zircaloy-3C, and resulting in Zircaloy-4 with tin addition (from 1.30% to 1.50%). Zircaloy-4 is now outdated for PWR and new zirconium alloys with niobium are used (M5, ZIRLO...) they present a better resistance to corrosion, to hydridation, to creep and they are less prone to dimensional changes under irradiation. (A.C.)

In situ Investigation of Oxide Films on Zirconium Alloy in PWR Primary Water Chemistry

Energy Technology Data Exchange (ETDEWEB)

Kim, Taeho; Choi, Kyoung Joon; Yoo, Seung Chang; Kim, Ji Hyun [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2015-05-15

Zirconium alloys are used as fuel cladding materials in nuclear power reactors, because these materials have a very low thermal neutron capture cross section as well as desirable mechanical properties. However, the Fukushima accident shows that the oxidation behavior of zirconium alloy is an important issue because the zirconium alloy functions as a shield of nuclear material (i.e., uranium, fission gas), and the degradation on zirconium cladding directly causes severe accident on nuclear power plant. Therefore, to ensure the safety of nuclear power reactors, the performance and sustainability of nuclear fuel should be understood. Currently, the water-metal interface is regarded as the rate-controlling site governing the rapid oxidation transition in high-burn-up fuels. Zirconium oxide is formed at the water-metal interface, and its structure and phase play an important role in determining its mechanical properties. In the early stage of the oxidation process, zirconium oxide with both tetragonal and monoclinic phases is formed. With an increase in the oxidation time to 150 h, the unstable tetragonal phase disappears and the monoclinic phase is dominant and possibly because of the stress relaxation according to previous and present results.

Synthesis, Characterization and Antimicrobial Activity of Zirconium (IV) Complexes

Energy Technology Data Exchange (ETDEWEB)

Sharma, Shobhana; Jain, Asha; Saxena, Sanjiv [Univ. of Rajasthan, Jaipur (India)

2012-08-15

Heteroleptic complexes of zirconium (IV) derived from bulky Schiff base ligands containing a sulphur atom and oximes of heterocyclic β-diketones of the general formula ZrLL' (where L'H{sub 2}=RCNH(C{sub 6}H{sub 4})SC : C(OH)N(C{sub 6}H{sub 5})N : CCH{sub 3}, R=-C{sub 6}H{sub 5}, -C{sub 6}H{sub 4}Cl(p) and L'H{sub 2}=R'C : (NOH)C : C(OH)N(C{sub 6}H{sub 5})N : CCH{sub 3}, R' = -CH{sub 2}CH{sub 3}, -C{sub 6}H{sub 5}, -C{sub 6}H{sub 4}Cl (p) were prepared by the reactions of zirconium tetrachloride with disodium salts of Schiff bases (L Na{sub 2}) and oximes of heterocyclic β-diketones (L' Na{sub 2}) in 1:1:1 molar ratio in dry refluxing THF. The structures of these monomeric zirconium (IV) complexes were elucidated with the help of elemental analysis, molecular weight measurements, spectroscopic (IR, NMR and mass) studies. A distorted trigonal bipyramidal geometry may be suggested for these heteroleptic zirconium (IV) complexes. The ligands (bulky Schiff base ligands containing a sulphur atom and oximes of heterocyclic β-diketones) and their heteroleptic complexes of zirconium (IV) were screened against A. flavus, P. aeruginesa and E. coli.

An evaluated neutronic data file for elemental zirconium

International Nuclear Information System (INIS)

Smith, A.B.; Chiba, S.

1994-09-01

A comprehensive evaluated neutronic data file for elemental zirconium is derived and presented in the ENDF/B-VI formats. The derivation is based upon measured microscopic nuclear data, augmented by model calculations as necessary. The primary objective is a quality contemporary file suitable for fission-reactor development extending from conventional thermal to fast and innovative systems. This new file is a significant improvement over previously available evaluated zirconium files, in part, as a consequence of extensive new experimental measurements reported elsewhere

Adsorption of zirconium from nitric acid solutions on hydrated tin dioxide

Energy Technology Data Exchange (ETDEWEB)

Tret' yakov, S Ya; Sharygin, L M; Egorov, Yu V

1977-01-01

Adsorption of zirconium from nitric acid solutions has been studied with the use of the labeled atom method on hydrated tin dioxide depending on the sorbate concentration, pH and prehistory of the solution. It has been found that adsorption behavior of zirconium essentially depends on its state in the solution.

Extending FEAST-METAL for analysis of low content minor actinide bearing and zirconium rich metallic fuels for sodium fast reactors

Energy Technology Data Exchange (ETDEWEB)

Karahan, Aydin, E-mail: karahan@mit.edu [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology, 77 Massachusetts Avenue, Cambridge MA 24-204 (United States)

2011-07-15

Computational models in FEAST-METAL fuel behaviour code have been upgraded to simulate minor actinide bearing zirconium rich metallic fuels for use in sodium fast reactors. Increasing the zirconium content to 20-40 wt.% causes significant changes in fuel slug microstructure affecting thermal, mechanical, chemical, and fission gas behaviour. Inclusion of zirconium rich phase reduces the fission gas swelling rate significantly in early irradiation. Above the threshold fission gas swelling, formation of micro-cracks, and open pores increase material compliancy enhance diffusivity, leading to rapid fuel gas swelling, interconnected porosity development and release of the fission gases and helium. Production and release of helium was modelled empirically as a function of americium content and fission gas production, consistent with previous Idaho National Laboratory studies. Predicted fuel constituent redistribution is much smaller compared to typical U-Pu-10Zr fuel operated at EBR-II. Material properties such as fuel thermal conductivity, modulus of elasticity, and thermal expansion coefficient have been approximated using the available database. Creep rate and fission gas diffusivity of high zirconium fuel is lowered by an order of magnitude with respect to the reference low zirconium fuel based on limited database and in order to match experimental observations. The new code is benchmarked against the AFC-1F fuel assembly post irradiation examination results. Satisfactory match was obtained for fission gas release and swelling behaviour. Finally, the study considers a comparison of fuel behaviour between high zirconium content minor actinide bearing fuel and typical U-15Pu-6Zr fuel pins with 75% smear density. The new fuel has much higher fissile content, allowing for operating at lower neutron flux level compared to fuel with lower fissile density. This feature allows the designer to reach a much higher burnup before reaching the cladding dose limit. On the other

Extending FEAST-METAL for analysis of low content minor actinide bearing and zirconium rich metallic fuels for sodium fast reactors

Science.gov (United States)

Karahan, Aydın

2011-07-01

Computational models in FEAST-METAL fuel behaviour code have been upgraded to simulate minor actinide bearing zirconium rich metallic fuels for use in sodium fast reactors. Increasing the zirconium content to 20-40 wt.% causes significant changes in fuel slug microstructure affecting thermal, mechanical, chemical, and fission gas behaviour. Inclusion of zirconium rich phase reduces the fission gas swelling rate significantly in early irradiation. Above the threshold fission gas swelling, formation of micro-cracks, and open pores increase material compliancy enhance diffusivity, leading to rapid fuel gas swelling, interconnected porosity development and release of the fission gases and helium. Production and release of helium was modelled empirically as a function of americium content and fission gas production, consistent with previous Idaho National Laboratory studies. Predicted fuel constituent redistribution is much smaller compared to typical U-Pu-10Zr fuel operated at EBR-II. Material properties such as fuel thermal conductivity, modulus of elasticity, and thermal expansion coefficient have been approximated using the available database. Creep rate and fission gas diffusivity of high zirconium fuel is lowered by an order of magnitude with respect to the reference low zirconium fuel based on limited database and in order to match experimental observations. The new code is benchmarked against the AFC-1F fuel assembly post irradiation examination results. Satisfactory match was obtained for fission gas release and swelling behaviour. Finally, the study considers a comparison of fuel behaviour between high zirconium content minor actinide bearing fuel and typical U-15Pu-6Zr fuel pins with 75% smear density. The new fuel has much higher fissile content, allowing for operating at lower neutron flux level compared to fuel with lower fissile density. This feature allows the designer to reach a much higher burnup before reaching the cladding dose limit. On the other

Extending FEAST-METAL for analysis of low content minor actinide bearing and zirconium rich metallic fuels for sodium fast reactors

International Nuclear Information System (INIS)

Karahan, Aydin

2011-01-01

Computational models in FEAST-METAL fuel behaviour code have been upgraded to simulate minor actinide bearing zirconium rich metallic fuels for use in sodium fast reactors. Increasing the zirconium content to 20-40 wt.% causes significant changes in fuel slug microstructure affecting thermal, mechanical, chemical, and fission gas behaviour. Inclusion of zirconium rich phase reduces the fission gas swelling rate significantly in early irradiation. Above the threshold fission gas swelling, formation of micro-cracks, and open pores increase material compliancy enhance diffusivity, leading to rapid fuel gas swelling, interconnected porosity development and release of the fission gases and helium. Production and release of helium was modelled empirically as a function of americium content and fission gas production, consistent with previous Idaho National Laboratory studies. Predicted fuel constituent redistribution is much smaller compared to typical U-Pu-10Zr fuel operated at EBR-II. Material properties such as fuel thermal conductivity, modulus of elasticity, and thermal expansion coefficient have been approximated using the available database. Creep rate and fission gas diffusivity of high zirconium fuel is lowered by an order of magnitude with respect to the reference low zirconium fuel based on limited database and in order to match experimental observations. The new code is benchmarked against the AFC-1F fuel assembly post irradiation examination results. Satisfactory match was obtained for fission gas release and swelling behaviour. Finally, the study considers a comparison of fuel behaviour between high zirconium content minor actinide bearing fuel and typical U-15Pu-6Zr fuel pins with 75% smear density. The new fuel has much higher fissile content, allowing for operating at lower neutron flux level compared to fuel with lower fissile density. This feature allows the designer to reach a much higher burnup before reaching the cladding dose limit. On the other

Study on the scattering law and scattering kernel of hydrogen in zirconium hydride

International Nuclear Information System (INIS)

Jiang Xinbiao; Chen Wei; Chen Da; Yin Banghua; Xie Zhongsheng

1999-01-01

The nuclear analytical model of calculating scattering law and scattering kernel for the uranium zirconium hybrid reactor is described. In the light of the acoustic and optic model of zirconium hydride, its frequency distribution function f(ω) is given and the scattering law of hydrogen in zirconium hydride is obtained by GASKET. The scattering kernel σ l (E 0 →E) of hydrogen bound in zirconium hydride is provided by the SMP code in the standard WIMS cross section library. Along with this library, WIMS is used to calculate the thermal neutron energy spectrum of fuel cell. The results are satisfied

Startup of Torrey Pines Mark III and Puerto Rico Nuclear Center reactors with TRIGA-FLIP fuel

Energy Technology Data Exchange (ETDEWEB)

Chesworth, R. H. [Gulf E and ES, San Diego, CA (United States)

1972-07-01

This paper discusses the characteristics of TRIGA FLIP cores in two different geometries: the normal TRIGA single-rod geometry as typified by the installation in the Torrey Pines Mark III reactor; and the four-rod cluster geometry as typified by the conversion core installed in the Puerto Rico Nuclear Center reactor at Mayaguez. In both reactors the fuel is 8-1/2 wt % uranium, 70% enriched in U-235. The hydrogen to zirconium atom ratio is 1.5 to 1.65 and the cladding material is stainless steel. The basic neutronic characteristics of the fuel in both reactor installations are briefly discussed.

Neutron-induced displacement damage analysis (with particular reference to zirconium)

International Nuclear Information System (INIS)

Woo, C.H.

1978-10-01

Neutron irradiation produces damage in a solid, initially in the form of atomic displacements. As a first step to understanding the effects of irradiation damage in reactor structural materials, information on the initial atomic displacements is necessary. The computer program DISPKAN, based on an extension and generalization of the program RICE, written at ORNL for such calculations, has been developed and installed on the CDC system at CRNL. Using neutron scattering data from ENDF/B files, DISPKAN performs a displacement and PKO analysis on the initial damage caused by neutrons from a given spectrum. The following quantities are calculated: the displacement rate per unit neutron flux, the PKO spectrum, the displacement spectrum, the fraction of PKO's with energy above T, the fraction of displacements produced by PKO's with energy above T, the average PKO energy, the average number of displacements produced per PKO, and the total number of PKO's produced per atom of the solid per unit fluence. The input and output formats of the program are explained. Sample runs are demonstrated. Results for zirconium, exposed to five neutron spectra typically available to experimentalist, are given to illustrate the spectral dependence of the initial displacement events. (author)

Sliding wear and friction behavior of zirconium alloy with heat-treated Inconel718

Energy Technology Data Exchange (ETDEWEB)

Kim, J.H., E-mail: kimjhoon@cnu.ac.kr [Dept. of Mechanical Design Engineering, Chungnam National University, 99 Daehak-ro, Yuseong-gu, Daejeon 305-764 (Korea, Republic of); Park, J.M. [Dept. of Mechanical Design Engineering, Chungnam National University, 99 Daehak-ro, Yuseong-gu, Daejeon 305-764 (Korea, Republic of); Park, J.K.; Jeon, K.L. [Nuclear Fuel Technology Department, Korea Nuclear Fuel, 1047 Daedukdae-ro, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

2014-04-01

In water-cooled nuclear reactors, the sliding of fuel rod can lead to severe wear and it is an important issue to sustain the structural integrity of nuclear reactor. In the present study, sliding wear behavior of zirconium alloy in dry and water environment using Pin-On-Disk sliding wear tester was investigated. Wear resistance of zirconium alloy against heat-treated Inconel718 pin was examined at room temperature. Sliding wear tests were carried out at different sliding distance, axial load and sliding speed based on ASTM (G99-05). The results of these experiments were verified with specific wear rate and coefficient of friction. The micro-mechanisms responsible for wear in zirconium alloy were identified to be microcutting and microcracking in dry environment. Moreover, micropitting and delamination were observed in water environment.

Effect of nitrogen flow ratio on structure and properties of zirconium ...

Indian Academy of Sciences (India)

Abstract. In this study, zirconium nitride thin films were deposited on Si substrates by ion beam sputtering (IBS). Influence of N2/(N2+Ar) on the structural and physical properties of the films has been investigated with respect to the atomic ratio between nitrogen and zirconium. It was found that the thickness of layers ...

Standard Specification for Nuclear Grade Zirconium Oxide Pellets

CERN Document Server

American Society for Testing and Materials. Philadelphia

2009-01-01

1.1 This specification applies to pellets of stabilized zirconium oxide used in nuclear reactors. 1.2 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.

Minimalistic Liquid-Assisted Route to Highly Crystalline α-Zirconium Phosphate.

Science.gov (United States)

Cheng, Yu; Wang, Xiaodong Tony; Jaenicke, Stephan; Chuah, Gaik-Khuan

2017-08-24

Zirconium phosphates have potential applications in areas of ion exchange, catalysis, photochemistry, and biotechnology. However, synthesis methodologies to form crystalline α-zirconium phosphate (Zr(HPO 4 ) 2 ⋅H 2 O) typically involve the use of excess phosphoric acid, addition of HF or oxalic acid and long reflux times or hydrothermal conditions. A minimalistic sustainable route to its synthesis has been developed by using only zirconium oxychloride and concentrated phosphoric acid to form highly crystalline α-zirconium phosphate within hours. The morphology can be changed from platelets to rod-shaped particles by fluoride addition. By varying the temperature and time, α-zirconium phosphate with particle sizes from nanometers to microns can be obtained. Key features of this minimal solvent synthesis are the excellent yields obtained with high atom economy under mild conditions and ease of scalability. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

Effects of titanium and zirconium on iron aluminide weldments

Energy Technology Data Exchange (ETDEWEB)

Mulac, B.L.; Edwards, G.R. [Colorado School of Mines, Golden, CO (United States). Center for Welding, Joining, and Coatings Research; Burt, R.P. [Alumax Technical Center, Golden, CO (United States); David, S.A. [Oak Ridge National Lab., TN (United States). Metals and Ceramics Div.

1997-12-01

When gas-tungsten arc welded, iron aluminides form a coarse fusion zone microstructure which is susceptible to hydrogen embrittlement. Titanium inoculation effectively refined the fusion zone microstructure in iron aluminide weldments, but the inoculated weldments had a reduced fracture strength despite the presence of a finer microstructure. The weldments fractured by transgranular cleavage which nucleated at cracked second phase particles. With titanium inoculation, second phase particles in the fusion zone changed shape and also became more concentrated at the grain boundaries, which increased the particle spacing in the fusion zone. The observed decrease in fracture strength with titanium inoculation was attributed to increased spacing of second phase particles in the fusion zone. Current research has focused on the weldability of zirconium- and carbon-alloyed iron aluminides. Preliminary work performed at Oak Ridge National Laboratory has shown that zirconium and carbon additions affect the weldability of the alloy as well as the mechanical properties and fracture behavior of the weldments. A sigmajig hot cracking test apparatus has been constructed and tested at Colorado School of Mines. Preliminary characterization of hot cracking of three zirconium- and carbon-alloyed iron aluminides, each containing a different total concentration of zirconium at a constant zirconium/carbon ratio of ten, is in progress. Future testing will include low zirconium alloys at zirconium/carbon ratios of five and one, as well as high zirconium alloys (1.5 to 2.0 atomic percent) at zirconium/carbon ratios of ten to forty.

Zirconium analysis. Impurities determination by spark mass specrometry

International Nuclear Information System (INIS)

Anon.

Determination of impurities in zirconium, suitable for atomic content greater than 10 -8 but particularly adapted for low contents. The method is quantitative only if a reference sample is available (metallic impurities) [fr

Atomic scale modeling of twinning in hexagonal metals: germination and migration of disconnections in zirconium, titanium and magnesium

International Nuclear Information System (INIS)

Mackain, Olivier

2017-01-01

We perform an atomic scale study of twinning in three hexagonal close packed metals: zirconium, titanium and magnesium. For that, we use two energetic models, an empirical potential suited for the study of zirconium and ab initio calculations in order to compare the three metals. The study of perfect twin boundaries shows that their surface energies are not relevant to predict which twin system is going to activate experimentally. We study nucleation and then migration of disconnections, that is to say interfacial dislocations whose glide along twin plane leads to twin thickening. We show that disconnections nucleation, rather than their migration, is the rate limiting step of twin thickening. In particular, by developing a coupling with elastic theory, we extract the core energies of each disconnection. With this method we prove that for a given twin system, the disconnection with the lowest formation energy is always the one compatible with twin mode observed experimentally, even if its elastic energy is higher. Finally, we construct the preliminary elements of a model considering the effect of stress on nucleation of disconnection. (author) [fr

Silica intercalated crystalline zirconium phosphate-type materials

NARCIS (Netherlands)

1988-01-01

The present invention relates to intercalated crystalline zirconium phosphate-types compositions wherein the interlayers of said composition have been intercalated with three-dimensional silicon oxide pillars whereby the pillars comprise at least two silicon atom layers parallel to the clay

Production kinetics of zirconium tetrachloride

International Nuclear Information System (INIS)

Sudjoko, D.; Masduki, B.; Sunardjo; Sulistyo, B.

1996-01-01

This research was intended to study the kinetics of zirconium tetrachloride production. The process was carried out in semi continuous reactor, equipped with heater, temperature controller, sublimator and scrubber. The variables investigated were time, temperature and the pellet forming pressure. Within the range of variables studied, the expression of the process in the chemical reaction controller region and diffusion controller region were both presented. (author)

Measurements of the effective total and resonance absorption cross sections for zircaloy-2 and zirconium

Energy Technology Data Exchange (ETDEWEB)

Kocic, A; Markovic, V [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1967-04-15

Zirconium and zircaloy-2 alloy, as constructive materials, have found wide application in reactor technology, especially in heavy water systems for two reasons: a) low neutron absorption cross section, b) good mechanical properties. The thickness of the zirconium and zircaloy-2 for different applications varies from several tenths of a millimeter to about ten millimeters. Therefore, to calculate reactor systems it is desirable to know the effective neutron absorption cross section for the range of thicknesses mention above. The thermal neutron cross sections for these materials are low and no appreciable variation of the effective neutron cross section occurs even for the largest thicknesses. However, this is not true for effective resonance absorption. On the other hand, due to the lack of detailed knowledge of the zirconium resonances, calculations of the effective resonance integrals cannot be performed. Therefore it is necessary to measure the effective total and resonance absorption cross section for zirconium (author)

Electrochemical impedance spectroscopic study of passive zirconium

Energy Technology Data Exchange (ETDEWEB)

Ai Jiahe; Chen Yingzi [Center for Electrochemical Science and Technology, Department of Materials Science and Engineering, Pennsylvania State University, University Park, PA 16802 (United States); Urquidi-Macdonald, Mirna [Department of Engineering Science and Mechanics, Pennsylvania State University, University Park, PA 16802 (United States); Macdonald, Digby D. [Center for Electrochemical Science and Technology, Department of Materials Science and Engineering, Pennsylvania State University, University Park, PA 16802 (United States)], E-mail: ddm2@psu.edu

2008-09-30

Spent, unreproccessed nuclear fuel is generally contained within the operational fuel sheathing fabricated from a zirconium alloy (Zircaloy 2, Zircaloy 4, or Zirlo) and is then stored in a swimming pool and/or dry storage facilities until permanent disposal in a licensed repository. During this period, which begins with irradiation of the fuel in the reactor during operation, the fuel sheathing is exposed to various, aggressive environments. The objective of the present study was to characterize the nature of the passive film that forms on pure zirconium in contact with an aqueous phase [0.1 M B(OH){sub 3} + 0.001 M LiOH, pH 6.94] at elevated temperatures (in this case, 250 deg. C), prior to storage, using electrochemical impedance spectroscopy (EIS) with the data being interpreted in terms of the point defect model (PDM). The results show that the corrosion resistance of zirconium in high temperature, de-aerated aqueous solutions is dominated by the outer layer. The extracted model parameter values can be used in deterministic models for predicting the accumulation of general corrosion damage to zirconium under a wide range of conditions that might exist in some repositories.

Reactor irradiation effect on the physical-mechanical properties of zirconium carbides and niobium carbides

International Nuclear Information System (INIS)

Andrievskij, R.A.; Vlasov, K.P.; Shevchenko, A.S.; Lanin, A.G.; Pritchin, S.A.; Klyushin, V.V.; Kurushin, S.P.; Maskaev, A.S.

1978-01-01

A study has been made of the effect of the reactor radiation by a flux of neutrons 1.5x10 20 n/cm 2 (E>=1 meV) at radiation temperatures of 150 and 1100 deg C on the physico-mechanical properties of carbides of zirconium and niobium and their equimolar hard solution. A difference has been discovered in the behaviour of the indicated carbides under the effect of radiation. Under the investigated conditions of radiation the density of zirconium carbide is being decreased, while in the niobium carbide no actual volumetric changes occur. The increase of the lattice period in ZrC is more significant than in NbC. The electric resistance of ZrC is also changed more significantly than in the case of NbC, while for the microhardness a reverse relationship is observed. Strength and elasticity modulus change insignificantly in both cases. Resistance to crack formation shows a higher reduction for ZrC than for NbC, while the thermal strength shows an approximately similar increase. The equimolar hard solution of ZrC and NbC behaves to great extent similar to ZrC, although the change in electric resistance reminds of NbC while thermal strength changes differently. The study of the microstructure of the specimens has shown that radiation causes a large number of etching patterns-dislocations in NbC which are almost absent in ZrC

Electroless deposition process for zirconium and zirconium alloys

Science.gov (United States)

Donaghy, Robert E.; Sherman, Anna H.

1981-01-01

A method is disclosed for preventing stress corrosion cracking or metal embrittlement of a zirconium or zirconium alloy container that is to be coated on the inside surface with a layer of a metal such as copper, a copper alloy, nickel, or iron and used for holding nuclear fuel material as a nuclear fuel element. The zirconium material is etched in an etchant solution, desmutted mechanically or ultrasonically, oxidized to form an oxide coating on the zirconium, cleaned in an aqueous alkaline cleaning solution, activated for electroless deposition of a metal layer and contacted with an electroless metal plating solution. This method provides a boundary layer of zirconium oxide between the zirconium container and the metal layer.

A computer model for hydride blister growth in zirconium alloys

International Nuclear Information System (INIS)

White, A.J.; Sawatzky, A.; Woo, C.H.

1985-06-01

The failure of a Zircaloy-2 pressure tube in the Pickering unit 2 reactor started at a series of zirconium hydride blisters on the outside of the pressure tube. These blisters resulted from the thermal diffusion of hydrogen to the cooler regions of the pressure tube. In this report the physics of thermal diffusion of hydrogen in zirconium is reviewed and a computer model for blister growth in two-dimensional Cartesian geometry is described. The model is used to show that the blister-growth rate in a two-phase zirconium/zirconium-hydride region does not depend on the initial hydrogen concentration nor on the hydrogen pick-up rate, and that for a fixed far-field temperature there is an optimum pressure-type/calandria-tube contact temperature for growing blisters. The model described here can also be used to study large-scale effects, such as hydrogen-depletion zones around hydride blisters

Studies in group IV organometallic chemistry XXX. Synthesis of compounds containing tin---titanium and tin---zirconium bonds

NARCIS (Netherlands)

Creemers, H.M.J.C.; Verbeek, F.; Noltes, J.G.

1968-01-01

Starting from the tetrakis(diethylamino) derivatives of titanium and zirconium and pheyltin hydrides six intermetalic compounds contianing up to nine tin and titanium(or zirconium) atoms have been obtained by hydrostannolysis type reactions.

The degradation of zirconium alloys in nuclear reactors - a review

International Nuclear Information System (INIS)

Lim, D.; Graham, N.A.

1986-01-01

This report presents the findings of a survey of available non-Canadian literature on the oxidation and hydriding of zirconium alloys. Much of the literature was found to address the Zircaloys, particularly when used as fuel cladding subjected to a radioactive and oxidizing environment. Hydriding of Zircaloys is mainly attributed to oxidation. The survey revealed that Zr-Nb alloys have been included in some investigations; however, data on the long-term degradation of Zr-2.5 wt% Nb, in particular, were scarce. The reviewed literature did not lead to conclusions regarding the potential for accelerated hydriding due to corrosion at crevices and/or second-phase particles, nor did it lead to conclusions as to the potential for a 'breakaway' in oxidation and hydrogen acquisition in long service life of Zr-Nb alloys. Specific information on service experience in U.S.S.R. power reactors could not be obtained; however, most of the information surveyed leads to the conclusion that fuel channels having Zr-2.5 wt% Nb pressure tubes should perform satisfactorily with respect to degradation from corrosion and hydriding provided they are installed correctly and are not operated under conditions that are far removed from those anticipated in design. 91 refs

Design of an Extractive Distillation Column for the Environmentally Benign Separation of Zirconium and Hafnium Tetrachloride for Nuclear Power Reactor Applications

Directory of Open Access Journals (Sweden)

Le Quang Minh

2015-09-01

Full Text Available Nuclear power with strengthened safety regulations continues to be used as an important resource in the world for managing atmospheric greenhouse gases and associated climate change. This study examined the environmentally benign separation of zirconium tetrachloride (ZrCl4 and hafnium tetrachloride (HfCl4 for nuclear power reactor applications through extractive distillation using a NaCl-KCl molten salt mixture. The vapor–liquid equilibrium behavior of ZrCl4 and HfCl4 over the molten salt system was correlated with Raoult’s law. The molten salt-based extractive distillation column was designed optimally using a rigorous commercial simulator for the feasible separation of ZrCl4 and HfCl4. The molten salt-based extractive distillation approach has many potential advantages for the commercial separation of ZrCl4 and HfCl4 compared to the conventional distillation because of its milder temperatures and pressure conditions, smaller number of required separation trays in the column, and lower energy requirement for separation, while still taking the advantage of environmentally benign feature by distillation. A heat-pump-assisted configuration was also explored to improve the energy efficiency of the extractive distillation process. The proposed enhanced configuration reduced the energy requirement drastically. Extractive distillation can be a promising option competing with the existing extraction-based separation process for zirconium purification for nuclear power reactor applications.

Enhanced low-temperature oxidation of zirconium alloys under irradiation

International Nuclear Information System (INIS)

Cox, B.; Fidleris, V.

1989-01-01

The linear growth of relatively thick (>300 nm) interference-colored oxide films on zirconium alloy specimens exposed in the Advanced Test Reactor (ATR) coolant at ≤55 o C was unexpected. Initial ideas were that this was a photoconduction effect. Experiments to study photoconduction in thin anodic zirconium oxide (ZrO 2 ) films in the laboratory were initiated to provide background data. It was found that, in the laboratory, provided a high electric field was maintained across the oxide during ultraviolet (UV) irradiation, enhanced growth of oxide occurred in the irradiated area. Similarly enhanced growth could be obtained on thin thermally formed oxide films that were immersed in an electrolyte with a high electric field superimposed. This enhanced growth was found to be caused by the development of porosity in the barrier oxide layer by an enhanced local dissolution and reprecipitation process during UV irradiation. Similar porosity was observed in the oxide films on the ATR specimens. Since it is not thought that a high electric field could have been present in this instance, localized dissolution of fast-neutron primary recoil tracks may be the operative mechanism. In all instances, the specimens attempt to maintain the normal barrier-layer oxide thickness, which causes the additional oxide growth. Similar mechanisms may have operated during the formation of thick loosely adherent, porous oxides in homogeneous reactor solutions under irradiation, and may be the cause of enhanced oxidation of zirconium alloys in high-temperature water-cooled reactors in some water chemistries. (author)

Modelling of Zirconium and Hafnium separation using continuous annular chromatography

International Nuclear Information System (INIS)

Moch-Setyadji; Endang Susiantini

2014-01-01

Nuclear degrees of zirconium in the form of a metal alloy is the main material for fuel cladding of NPP. Zirconium is also used as sheathing UO 2 kernel in the form of ZrC as a substitute of SiC in the fuel elements of High Temperature Reactor (HTR). Difficulty separating hafnium from zirconium because it has a lot of similarities in the chemical properties of Zr and Hf. Annular chromatography is a device that can be used for separating of zirconium and hafnium to obtain zirconium nuclear grade. Therefore, it is necessary to construct the mathematical modelling that can describe the separation of zirconium and hafnium in the annular chromatography containing anion resin dowex-1X8. The aim of research is to perform separation simulation by using the equilibrium model and mass transfer coefficient resulted from research. Zr and Hf feed used in this research were 26 and 1 g/l, respectively. Height of resin (L), angular velocity (ω) and the superficial flow rate (uz) was varied to determine the effect of each parameter on the separation of Zr and Hf. By using Kd and Dv values resulted previous research. Simulation results showed that zirconium and hafnium can be separated using a continuous annular chromatography with high resin (long bed) 50 cm, superficial flow rate of 0.001 cm/s, the rotation speed of 0.006 rad/min and 20 cm diameter annular. In these conditions the results obtained zirconium concentration of 10,303.226 g/m 3 and hafnium concentration of 12.324 g/m 3 (ppm). (author)

Layer-by-layer deposition of zirconium oxide films from aqueous solutions for friction reduction in silicon-based microelectromechanical system devices

International Nuclear Information System (INIS)

Liu Junfu; Nistorica, Corina; Gory, Igor; Skidmore, George; Mantiziba, Fadziso M.; Gnade, Bruce E.

2005-01-01

This work reports layer-by-layer deposition of zirconium oxide on a Si surface from aqueous solutions using the successive ionic layer adsorption and reaction technique. The process consists of repeated cycles of adsorption of zirconium precursors, water rinse, and hydrolysis. The film composition was determined by X-ray photoelectron spectroscopy. The film thickness was determined by Rutherford backscattering spectrometry, by measuring the Zr atom concentration. The average deposition rate from a 0.1 M Zr(SO 4 ) 2 solution on a SiO 2 /Si surface is 0.62 nm per cycle. Increasing the acidity of the zirconium precursor solution inhibits the deposition of the zirconium oxide film. Atomic force microscopy shows that the zirconium oxide film consists of nanoparticles of 10-50 nm in the lateral dimension. The surface roughness increased with increasing number of deposition cycles. Friction measurements made with a microelectromechanical system device reveal a reduction of 45% in the friction coefficient of zirconium oxide-coated surfaces vs. uncoated surfaces in air

Irradiation effects of the zirconium oxidation and the uranium diffusion in zirconia; Effets d'irradiation sur l'oxydation du zirconium et la diffusion de l'uranium dans la zircone

Energy Technology Data Exchange (ETDEWEB)

Bererd, N

2003-07-01

The context of this study is the direct storage of spent fuel assemblies after operation in reactor. In order to obtain data on the capacities of the can as the uranium diffusion barrier, a fundamental study has been carried out for modelling the internal cladding surface under and without irradiation. The behaviour of zirconium in reactor conditions has at first been studied. A thin uranium target enriched with fissile isotope has been put on a zirconium sample, the set being irradiated by a thermal neutrons flux leading to the fission of the deposited uranium. The energetic history of the formed fission products has revealed two steps: 1)the zirconium oxidation and 2)the diffusion of uranium in the zirconia formed at 480 degrees C. A diffusion coefficient under irradiation has been measured. Its value is 10{sup -15} cm{sup 2}.s{sup -1}. In order to be able to reveal clearly the effect of the irradiation by the fission products on the zirconium oxidation, measurements of thermal oxidation and under {sup 129}Xe irradiation have been carried out. They have shown that the oxidation is strongly accelerated by the irradiation and that the temperature is negligible until 480 degrees C. On the other hand, the thermal diffusion of the uranium in zirconium and in zirconia has been studied by coupling ion implantation and Rutherford backscattering spectroscopy. This study shows that the uranium diffuses in zirconium and is trapped in zirconia in a UO{sub 3} shape. (O.M.)

Methods for the preparation of ultra-pure anhydrous zirconium tetrafluoride from zirconium tetraborohydride, researches in connection with halide glasses

International Nuclear Information System (INIS)

Tortevois, R.

1990-01-01

The synthesis of ultrapure zirconium tetrafluoride, the main component of fluorozirconate based optical fibers, was successfully attempted from zirconium tetraborohydride. Of the fluorinating agents used, nitrogen trifluoride doesn't react with zirconium tetraborohydride while xenon difluoride reacts too violently and leads to phases which contain boron. The fluorination in a compatible solvent enabled us to minimize the degradation. The best results were obtained with the fluorination of Zr(BH 4 ) 4 dissolved in CFCl 3 at -40 deg C by anhydrous HF. Using several analytical methods such as graphite furnace atomic absorption and proton activation, we analyzed the purity. The degree of transition element impurities is less than the ppm level for ZrF 4 . The dehydration of ZrF 4 ,H 2 O and ZrF 4 ,3H 2 O at room temperature by CIF 3 in gaseous and liquid state was also investigated. At exceptionally low temperature, this process allows oxide and oxyfluoride components to be reduced

Design basis for creep of zirconium alloy components in a fast neutron flux

International Nuclear Information System (INIS)

Ross-Ross, P.A.; Fidleris, V.

1975-01-01

The chalk River Nuclear Laboratory's experience with the creep of zirconium alloys in a neutron flux is described. Fast neutron flux changes the creep behaviour of zirconium alloys and new design criteria for in-reactor applications are needed. From experimental results empirical relations describing the effects of neutron flux, stress, temperature, time and anisotropy on creep rate were established. The relations are applied to the design of pressure tubes. (author)

Design basis for creep of zirconium alloy components in a fast neutron flux

International Nuclear Information System (INIS)

Ross-Ross, P.A.; Fidleris, V.

1974-01-01

The Chalk River Nuclear Laboratory's experience with the creep of zirconium alloys in a neutron flux is described. Fast neutron flux changes the creep behavior of zirconium alloys and new design criteria for in-reactor applications are needed. From experimental results empirical relations describing the effects of neutron flux, stress, temperature, time, and anisotropy on creep rate were established. The relations are applied to the design of pressure tubes. (author)

Concept Feasibility Report for Electroplating Zirconium onto Uranium Foil - Year 2

Energy Technology Data Exchange (ETDEWEB)

Coffey, Greg W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Meinhardt, Kerry D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Joshi, Vineet V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pederson, Larry R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Burkes, Douglas [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2015-03-01

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

Hydrogen determination in magnesium, zirconium, sodium and lithium using installation, C2532

International Nuclear Information System (INIS)

Malikova, E.D.; Velyukhanov, V.P.; Makhinova, L.O.; Kunin, L.L.

1980-01-01

Techniques of hydrogen determination in magnesium, lithium, sodium and zirconium using the S 2532 installation are developed. The method of oxidizing melting using lead borate has been used for hydrogen determination in lithium and sodium and the method of vacuum extraction - for hydrogen determination in zirconium and magnesium. Zr and Mg extraction has been carried out in steel reactor at the temperatures of 1000 and 650 deg C, the time of extraction being 30 and 10 minutes respectively. A quartz reactor, temperatures of oxidizing melting of 700-800 deg C, the time of analysis 10 and 20 minutes have been used for sodium and lithium. A possibility to determine volumetric content of hydrogen in magnesium at the existing surface contaminations with hydrogen-containing compounds is shown [ru

Long-time corrosion and high-temperature oxidation of zirconium alloys applied on NPP like fuel elements cover

International Nuclear Information System (INIS)

Vrtilkova, V.; Novotny, L.; Lingart, S.; Doukha, R.; Yarosh, Ya.; Kolenchik, Ya.

2007-01-01

Zirconium is applying in nuclear energy since 50-th of last century in capacity of material for cover production for fuel elements, reactor fuel and structural parts, and mainly due to both corrosion stability and low effective cross section for thermal neutrons capture. Impurities in doping elements form and alloy production technology has influence on mechanical and corrosion properties of finite alloy. Long-time corrosion tests for several zirconium alloys in forcing autoclave under different reaction conditions were carried out. After that process kinetics was studied, mass increase, hydrogen formation, zirconium hydride forming morphology, zirconium oxide layer thickness have been determined as well

International strategic minerals inventory summary report; zirconium

Science.gov (United States)

Towner, R.R.

1992-01-01

Zircon, a zirconium silicate, is currently the most important commercial zirconium-bearing mineral. Baddeleyite, a natural form of zirconia, is less important but has some specific end uses. Both zircon and baddeleyite occur in hard-rock and placer deposits, but at present all zircon production is from placer deposits. Most baddeleyite production is from hard-rock deposits, principally as a byproduct of copper and phosphate-rock mining. World zirconium resources in identified, economically exploitable deposits are about 46 times current production rates. Of these resources, some 71 percent are in South Africa, Australia, and the United States. The principal end uses of zirconium minerals are in ceramic applications and as refractories, abrasives, and mold linings in foundries. A minor amount, mainly of zircon, is used for the production of hafnium-free zirconium metal, which is used principally for sheathing fuel elements in nuclear reactors and in the chemical-processing industry, aerospace engineering, and electronics. Australia and South Africa are the largest zircon producers and account for more than 70 percent of world output; the United States and the Soviet Union account for another 20 percent. South Africa accounts for almost all the world's production of baddeleyite, which is about 2 percent of world production of contained zirconia. Australia and South Africa are the largest exporters of zircon. Unless major new deposits are developed in countries that have not traditionally produced zircon, the pattern of world production is unlikely to change by 2020. The proportions, however, of production that come from existing producing countries may change somewhat.

Preparation and characterization of very pure zirconium tetrafluoride. Application to fluorinated glass

International Nuclear Information System (INIS)

Bridenne, M.

1986-12-01

The synthesis of anhydrous and very pure zirconium tetrafluoride from zirconium tetraborohydride is studied. Zr F 4 is used for fabrication of fluorozirconate glass. Zr (BH 4 ) 4 is purified by sublimation. Two fluorinating agents F 2 and anhydrous HF are used for fluorination. The apparatus is made of fluorinated polymers and a Kel-F prototype reactor was realized. 20 g of Zr F 4 are obtained in 44 hrs with a yield of 88 %. Purity is characterized by chemical analysis (atomique absorption spectroscopy and spark mass spectroscopy) and absorption of an optical fiber made of zirconium tetrafluoride. Cr, Ni, Co and Cu content is lower than 0.1 ppm. Possibility of pilot scale production is discussed [fr

Mechanical and irradiation properties of zirconium alloys irradiated in HANARO

International Nuclear Information System (INIS)

Kwon, Oh Hyun; Eom, Kyong Bo; Kim, Jae Ik; Suh, Jung Min; Jeon, Kyeong Lak

2011-01-01

These experimental studies are carried out to build a database for analyzing fuel performance in nuclear power plants. In particular, this study focuses on the mechanical and irradiation properties of three kinds of zirconium alloy (Alloy A, Alloy B and Alloy C) irradiated in the HANARO (High-flux Advanced Neutron Application Reactor), one of the leading multipurpose research reactors in the world. Yield strength and ultimate tensile strength were measured to determine the mechanical properties before and after irradiation, while irradiation growth was measured for the irradiation properties. The samples for irradiation testing are classified by texture. For the irradiation condition, all samples were wrapped into the capsule (07M-13N) and irradiated in the HANARO for about 100 days (E > 1.0 MeV, 1.1 10 21 n/cm 2 ). These tests and results indicate that the mechanical properties of zirconium alloys are similar whether unirradiated or irradiated. Alloy B has shown the highest yield strength and tensile strength properties compared to other alloys in irradiated condition. Even though each of the zirconium alloys has a different alloying content, this content does not seem to affect the mechanical properties under an unirradiated condition and low fluence. And all the alloys have shown the tendency to increase in yield strength and ultimate tensile strength. Transverse specimens of each of the zirconium alloys have a slightly lower irradiation growth tendency than longitudinal specimens. However, for clear analysis of texture effects, further testing under higher irradiation conditions is needed

Thermal Energetic Reactor with High Reproduction of Fission Materials

International Nuclear Information System (INIS)

Kotov, V.M.

2012-01-01

Existing thermal reactors are energy production scale limited because of low portion of raw uranium usage. Fast reactors are limited by reprocessing need of huge mass of raw uranium at the initial stage of development. The possibility of development of thermal reactors with high fission materials reproduction, which solves the problem, is discussed here. Neutron losses are decreased, uranium-thorium fuel with artificial fission materials equilibrium regime is used, additional in-core and out-core neutron sources are used for supplying of high fission materials reproduction. Liquid salt reactors can use dynamic loading regime for this purpose. Preferable construction is channel type reactor with heavy water moderator. Good materials for fuel element shells and channel walls are zirconium alloys enriched by 90Zr. Water cooled reactors with usage 12% of raw uranium and liquid metal cooled reactors with usage 25% of raw uranium are discussed. Reactors with additional neutron sources obtain full usage of raw uranium with small additional energy expenses. On the base of thermal reactors with high fission materials reproduction world atomic power engineering development supplying higher power and requiring smaller speed of raw uranium mining, than in the variant with fast reactors, is possible.

Self-interstitial atom clusters as obstacles to glide of 1/3{11-bar 00} edge dislocations in α-zirconium

International Nuclear Information System (INIS)

Voskoboynikov, R.E.; Osetsky, Yu.N.; Bacon, D.J.

2005-01-01

Atomic-scale details of interaction of a 1/3 {11-bar 00} edge dislocation with clusters of self-interstitial atoms (SIAs) in α-zirconium has been studied by computer simulation. Four typical clusters are considered. A triangular cluster of five SIAs lying within a basal plane bisected by the dislocation glide plane is not absorbed by the dislocation but acts as a moderately strong obstacle. A 3-D SIA cluster lying across the glide plane is completely absorbed by the dislocation by creation of super-jogs, and is a weak obstacle. Interaction of the dislocation with glissile SIA loops with perfect Burgers vector inclined at 60 deg. to the dislocation glide plane shows that the process depends on the vector orientation. Defects of the two orientations are strong obstacles, and one, which initially forms a sessile segment on the dislocation line, is particularly so

Hydrogen desorption kinetics from zirconium hydride and zirconium metal in vacuum

International Nuclear Information System (INIS)

Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.

2014-01-01

The kinetics of hydrogen desorption from zirconium hydride is important in many nuclear design and safety applications. In this paper, a coordinated experimental and modeling study has been used to explicitly demonstrate the applicability of existing kinetic theories for hydrogen desorption from zirconium hydride and α-zirconium. A static synthesis method was used to produce δ-zirconium hydride, and the crystallographic phases of the zirconium hydride were confirmed by X-ray diffraction (XRD). Three obvious stages, involving δ-zirconium hydride, a two-phase region, and α-zirconium, were observed in the hydrogen desorption spectra of two zirconium hydride specimens with H/Zr ratios of 1.62 and 1.64, respectively, which were obtained using thermal desorption spectroscopy (TDS). A continuous, one-dimensional, two-phase moving boundary model, coupled with the zero- and second-order kinetics of hydrogen desorption from δ-zirconium hydride and α-zirconium, respectively, has been developed to reproduce the TDS experimental results. A comparison of the modeling predictions with the experimental results indicates that a zero-order kinetic model is valid for description of hydrogen flux away from the δ-hydride phase, and that a second-order kinetic model works well for hydrogen desorption from α-Zr if the activation energy of desorption is optimized to be 70% of the value reported in the literature

Quantification of chlorine in zirconium oxide and biological samples by instrumental NAA utilizing PCF of Dhruva reactor

International Nuclear Information System (INIS)

Shinde, Amol D.; Reddy, A.V.R.; Acharya, R.; Balaji Rao, Y.

2012-01-01

Recently studies on chlorine contents in various samples are being pursued due to its corrosive nature. Chlorine present at trace level in various finished products as well as powder is used as a raw material for production of different types of zircaloys used as structural materials in nuclear technology. As a part of quality assurance program, it is necessary to quantify chlorine accurately with suitable and simple technique. In the present work we have applied instrumental neutron activation analysis (INAA) utilizing its short-lived activation product ( 38 Cl, 37 min, 1642 and 2168 keV) for its estimation. Pneumatic Carrier Facility (PCF) of Dhruva reactor, BARC was used sample irradiation of zirconium oxide dry powder, synthetic wax and IAEA RMs 1515 (Apple leaves) and Lichen 336. (author)

The 10 MW multipurpose TRIGA reactor at Ongkharak Nuclear Research Center, Thailand

International Nuclear Information System (INIS)

Thurgood, B.E.; Razvi, J.; Whittemore, J.L.; Bhadrakom, K.

1997-01-01

General Atomics (GA), has been selected to lead a team of firms from the United States, Japan, Australia and Thailand to design, build and commission the Ongkharak Nuclear Research Center near Bangkok, Thailand, for the Office of Atomic Energy for Peace. The facilities to be provided comprise of: A Reactor Island, consisting of a 10 MW TRIGA reactor that takes full advantage of the inherent safety characteristics of uranium-zirconium hydride (UZrH) fuel; An Isotope Production Facility for the production of radioisotopes and radiopharmaceuticals using the TRIGA reactor; A Waste Processing and Storage Facility for the processing and storage of radioactive waste from the facility as well as other locations in Thailand. The centerpiece of the Center will be the TRIGA reactor, fueled with low-enriched UZrH fuel, cooled and moderated by light water, and reflected by beryllium and heavy water. The UZrH fueled reactor will have a rated steady state thermal power output of 10 MW, and will be capable of performing the following: Radioisotope production for medical, industrial and agricultural uses; Neutron transmutation doping of silicon; Beam experiments such as Neutron Scattering, Neutron Radiography (NR), and Prompt Gamma Neutron Activation Analysis (PGNAA); Medical therapy of patients using Boron Neutron Capture Therapy (BNCT); Applied research and technology development in the nuclear field; Training in principles of reactor operation, reactor physics, reactor experiments, etc. (author)

Multiscale modelling of hydrogen embrittlement in zirconium alloys

Energy Technology Data Exchange (ETDEWEB)

Majevadia, Jassel; Wenman, Mark; Balint, Daniel; Sutton, Adrian [Imperial College London (United Kingdom); Nazarov, Roman [MPIE, Dusseldorf (Germany)

2013-07-01

Delayed Hydride Cracking (DHC) is a commonly occurring embrittlement phenomenon in zirconium alloy fuel cladding within Pressurized Water Reactors (PWRs). DHC is caused by the accumulation of hydrogen atoms taken up by the metal, and the formation of brittle hydrides in the vicinity of crack tips. The rate of crack growth is limited by the rate of hydrogen diffusion to the crack, which can be modelled by solving a stress driven diffusion equation that incorporates the elastic interaction between defects. This of interest in the present work. The elastic interaction is calculated by combining defect forces determined through Density Functional Theory (DFT) simulations, and an exact solution for the anisotropic elastic field of an edge dislocation in Zr. making it possible to determine the interaction energy without the need to simulate directly a hydrogen atom in the presence of a crack or dislocation, which is computationally prohibitive with DFT. The result of the elastic interaction energy calculations can be utilised to determine the segregation of hydrogen to a crack tip for varying crack tip geometries, and in the presence of other crystal defects. This is done by implementing a diffusion equation for hydrogen within a discrete dislocation dynamics simulation. In the present work a model has been developed to demonstrate the effect of a single dislocation on hydrogen diffusion to create a Cottrell atmosphere.

Reactor core structure

International Nuclear Information System (INIS)

Higashinakagawa, Emiko; Sato, Kanemitsu.

1992-01-01

Taking notice on the fact that Fe based alloys and Ni based alloys are corrosion resistant in a special atmosphere of a nuclear reactor, Fe or Ni based alloys are applied to reactor core structural components such as fuel cladding tubes, fuel channels, spacers, etc. On the other hand, the neutron absorption cross section of zirconium is 0.18 barn while that of iron is 2.52 barn and that of nickel is 4.6 barn, which amounts to 14 to 25 times compared with that of zirconium. Accordingly, if the reactor core structural components are constituted by the Fe or Ni based alloys, neutron economy is lowered. Since it is desirable that neutrons contribute to uranium fission with least absorption to the reactor core structural components, the reactor core structural components are constituted with the Fe or Ni based alloys of good corrosion resistance only at a portion in contact with reactor water, that is, at a surface portion, while the main body is constituted with zircalloy in the present invention. Accordingly, corrosion resistnace can be kept while keeping small neutron absorption cross section. (T.M.)

A density functional theory study of a silica-supported zirconium monohydride catalyst for depolymerization of polyethylene

Energy Technology Data Exchange (ETDEWEB)

Mortensen, J.J.; Parrinello, M.

2000-04-06

A silica-supported zirconium hydride catalyst for depolymerization of polyethylene is studied using density functional theory (DFT) together with a generalized gradient approximation (GGA) for the exchange and correlation energy. The (100) and (111) surfaces of {beta}-cristobalite are used as two possible models of a silica surface. Based on the experimental surface structure determined by J. Corker et al., they propose a detailed atomic model of the zirconium monohydride that is believed to be the active site for depolymerization of polyolefins. The model of the zirconium monohydride on the (100) surface is found to be very stable and the structure is in good agreement with extended X-ray absorption fine structure (EXAFS) measurements. Depolymerization of a small polyolefin chain (C{sub 3}H{sub 8}) was carried out to give CH{sub 4} and C{sub 2}H{sub 6} by addition of H{sub 2}. The rate-limiting step is a {beta}-methyl transfer to the zirconium atom, and the activation energy is 29 kcal/mol on the (100) surface.

Deformation mechanisms and irradiation effects in zirconium alloys. A multi-scale study

International Nuclear Information System (INIS)

Onimus, Fabien

2015-01-01

Zirconium alloys have been used for more than 30 years in the nuclear industry as structural materials for the fuel assemblies of pressurized water reactors. In particular, the cladding tube, made of zirconium alloys, constitutes the first barrier against the dissemination of radioactive elements. It is therefore essential to have a good understanding and prediction of the mechanical behavior of these materials in various conditions. The work presented in this dissertation deals with an experimental study and numerical simulations, at several length scales, of the deformation mechanisms and the mechanical behavior of zirconium alloys before irradiation, but also after irradiation and under irradiation. The mechanical behavior of zirconium single crystal has been determined, during an original study, using tensile test specimens containing large grains. Based on this study, crystal plasticity constitutive laws have been proposed. A polycrystalline model has also been developed to simulate the behavior of unirradiated zirconium alloys. A thorough Transmission Electron Microscopy (TEM) study has been able to clarify the deformation mechanisms of zirconium alloys occurring after irradiation. The clearing of loops by gliding dislocations leading to the dislocation channeling mechanism has been studied in details. This phenomenon has also been simulated using a dislocation dynamics code. The macroscopic consequences of this process have also been analyzed. A polycrystalline model taking into account the specificity of this mechanism has eventually been proposed. This approach has then been extended to the post-irradiation creep behavior. The recovery of radiation defects during creep tests has been characterized by TEM and modeled using cluster dynamics method. Deformation modes during creep have also been studied and a simple model for the creep behavior has eventually been proposed. Finally, the mechanism responsible for the acceleration of irradiation growth that

Radiochemical determination of zirconium by inductively coupled plasma mass spectrometry (ICPMS)

International Nuclear Information System (INIS)

Oliveira, Thiago C.; Oliveira, Arno Heeren de

2013-01-01

The zirconium isotope 93 Zr is a long-lived pure β-particle-emitting radionuclide thus occurring as one of the radionuclides found in nuclear reactors. It's produced from 235 U fission and from 92 Zr neutron activation. Due to its long half-life, 93 Zr is one of the interest radionuclides for assessment studies performance of waste storage or disposal. Measurement of 93 Zr is difficult owing to its trace level concentration and its low activity in nuclear wastes and further because its certified standards are not frequently available. The aim of this work was to apply a selective radiochemical separation methodology for 93 Zr determination in nuclear waste and analyze it by Inductively Coupled Plasma Mass Spectrometry (ICPMS). To set up the zirconium radiochemical separation procedure, a zirconium tracer solution was used in order to follow the zirconium behavior during the radiochemical separation. A tracer solution containing the main interferences, Ba, Co, Eu, Fe, Mn, Nb, Ni, Sr, and Y was used in order to verify the decontamination factor during separation process. The limit of detection of 0,039 ppb was obtained for zirconium standard solutions by ICPMS. Then, the protocol will be applied to low level waste (LLW) and intermediate level waste (ILW) from nuclear power plants. (author)

Anelastic relaxation peaks in single crystals of zirconium-oxygen alloys

International Nuclear Information System (INIS)

Ritchie, I.G.; Sprungmann, K.W.; Atrens, A.; Rosinger, H.E.; CEA Centre d'Etudes Nucleaires de Grenoble, 38

1977-01-01

Relaxations of the compliances S 11 -S 12 and S 44 have been observed in single crystals of zirconium-oxygen alloys tested in flexure and in torsion respectively. The relaxations are attributed to the stress-induced reorientation of substitutional impurity atoms (s) paired with interstitial oxygen atoms (i). The results demonstrate that the jump of the interstitial parallel to the basal plane dominates in the reorientation of the s-i pair

Precipitation of γ-zirconium hydride in zirconium

International Nuclear Information System (INIS)

Carpenter, G.J.C.

1978-01-01

A mechanism for the precipitation of γ-zirconium hydride in zirconium is presented which does not require the diffusion of zirconium. The transformation is completed by shears caused by 1/3 (10 anti 10) Shockley partial dislocations on alternate zirconium basal planes, either by homogeneous nucleation or at lattice imperfections. Homogeneous nucleation is considered least likely in view of the large nucleation barrier involved. Hydrides may form at dislocations by the generation of partials by means of either a pole or ratchet mechanism. The former requires dislocations with a component of Burgers vector along the c-axis, but contrast experiments show that these are not normally observed in annealed zirconium. It is therefore most likely that intragranular hydrides form at the regular 1/3 (11 anti 20) dislocations, possibly by means of a ratchet mechanism. Contrast experiments in the electron microscope show that the precipitates have a shear character consistent with the mechanism suggested. The possibility that the shear dislocations associated with the hydrides are emissary dislocations is considered and a model suggested in which this function is satisfied together with the partial relief of misfit stresses. The large shear strains associated with the precipitation mechanism may play an important role in the preferential orientation of hydrides under stress

Determination of hafnium at the 10−4% level (relative to zirconium content) using neutron activation analysis, inductively coupled plasma mass spectrometry and inductively coupled plasma atomic emission spectrometry

International Nuclear Information System (INIS)

Smolik, Marek; Polkowska-Motrenko, Halina; Hubicki, Zbigniew; Jakóbik-Kolon, Agata; Danko, Bożena

2014-01-01

Graphical abstract: -- Highlights: •We worked out ICP-MS method of Hf determination in Zr and Zr compounds. •We used NAA method as reference one. •We obtained pure zirconium matrix by ion exchange (Diphonix ® resin). •These permit to determine ≥1 × 10 −4 % Hf in Zr sample by ICP MS with good precision and accuracy. -- Abstract: Hafnium at the very low level of 1–8 ppm (in relation to zirconium) was determined in zirconium sulfate solutions (originating from investigations of the separation of ca. 44 ppm Hf from zirconium by means of the ion exchange method) by using three independent methods: inductively coupled plasma mass spectrometry (ICP MS), neutron activation analysis (NAA) and inductively coupled plasma atomic emission spectrometry (ICP-AES). The results of NAA and ICP MS determinations were consistent with each other across the entire investigated range (the RSD of both methods did not exceed 38%). The results of ICP-AES determination were more diverse, particularly at less than 5 ppm Hf (RSD was significantly higher: 29–253%). The ion exchange method exploiting Diphonix ® resin proved sufficient efficiency in Zr–Hf separation when the initial concentration ratio of the elements ([Zr] 0 /[Hf] 0 ) ranged from 1200 to ca. 143,000

Standard specification for nuclear-grade zirconium oxide powder

CERN Document Server

American Society for Testing and Materials. Philadelphia

2008-01-01

1.1 This specification defines the physical and chemical requirements for zirconium oxide powder intended for fabrication into shapes, either entirely or partially of zirconia, for use in a nuclear reactor core. 1.2 The material described herein shall be particulate in nature. 1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.

Waterside corrosion of zirconium alloys in nuclear power plants

International Nuclear Information System (INIS)

Jeong, Yong Hwan; Baek, B. J.; Park, S. Y. and others

1999-08-01

The overview of corrosion and hydriding behaviors of Zr-based alloy under the conditions of the in-reactor service and in the absence of irradiation is introduced in this report. The metallurgical characteristics of Zr-based alloys and the thermo-mechanical treatments on the microstructures and the textures in the manufacturing process for fuel cladding are also introduced. The factors affecting the corrosion of Zr alloy in reactor are summarized. And the corrosion mechanism and hydrogen up-take are discussed based on the laboratory and in-reactor results. The phenomenological observations of zirconium alloy corrosion in reactors are summarized and the models of in-reactor corrosion are exclusively discussed. Finally, the effects of irradiation on the corrosion process in Zr alloy were investigated mainly based on the literature data. (author). 538 refs., 26 tabs., 105 figs

Shadow corrosion evaluation in the Studsvik R2 reactor

International Nuclear Information System (INIS)

Sanders, Ch.; Lysell, G.

2000-01-01

Post-irradiation examination has shown that increased corrosion occurs when zirconium alloys are in contact with or in proximity to other metallic objects. The observations indicate an influence of irradiation from the adjacent component as the enhanced corrosion occurs as a 'shadow' of the metallic object on the zirconium surface. This phenomenon could ultimately limit the lifetime of certain zirconium alloy components in the reactor. The Studsvik R2 materials test reactor has an In-Core Autoclave (INCA) test facility especially designed for water chemistry and materials research. The INCA facility has been evaluated and found suitable for shadow corrosion studies. The R2 reactor core containing the INCA facility was modeled with the Monte Carlo N-Particle (MCNP) code in order to evaluate the electron deposition in various materials and to develop a hypothesis of the shadow corrosion mechanism. (authors)

Fabrication of particulate metal fuel for fast burner reactors

International Nuclear Information System (INIS)

Ryu, Ho Jin; Lee, Sun Yong; Kim, Jong Hwan; Woo, Yoon Myung; Ko, Young Mo; Kim, Ki Hwan; Park, Jong Man; Lee, Chan Bok

2012-01-01

U Zr metallic fuel for sodium cooled fast reactors is now being developed by KAERI as a national R and D program of Korea. In order to recycle transuranic elements (TRU) retained in spent nuclear fuel, remote fabrication capability in a shielded hot cell should be prepared. Moreover, generation of long lived radioactive wastes and loss of volatile species should be minimized during the recycled fuel fabrication step. Therefore, innovative fuel concepts should be developed to address the fabrication challenges pertaining to TRU while maintaining good performances of metallic fuel. Particulate fuel concepts have already been proposed and tested at several experimental fast reactor systems and vipac ceramic fuel of RIAR, Russia is one of the examples. However, much less work has been reported for particulate metallic fuel development. Spherical uranium alloy particles with various diameters can be easily produced by the centrifugal atomization technique developed by KAERI. Using the atomized uranium and uranium zirconium alloy particles, we fabricated various kinds of powder pack, powder compacts and sintered pellets. The microstructures and properties of the powder pack and pellets are presented

Quantification of the distribution of hydrogen by nuclear microprobe at the Laboratory Pierre Sue in the width of zirconium alloy fuel clad of PWR reactors

International Nuclear Information System (INIS)

Raepsaet, C.; Bossis, Ph.; Hamon, D.; Bechade, J.L.; Brachet, J.C.

2007-01-01

Among the analysis techniques by ions beams, the micro ERDA (Elastic Detection Analysis) is an interesting technique which allows the quantitative distribution of the hydrogen in materials. In particular, this analysis has been used for hydride zirconium alloys, with the nuclear microprobe of the Laboratory Pierre Sue. This probe allows the characterization of radioactive materials. The technique principles are recalled and then two examples are provided to illustrate the fuel clad behavior in PWR reactors. (A.L.B.)

Lithium uptake and the corrosion of zirconium alloys in aqueous lithium hydroxide solutions

International Nuclear Information System (INIS)

Ramasubramanian, N.

1991-01-01

This paper reports on corrosion films on zirconium alloys that were analyzed for lithium by Atomic Absorption Spectroscopy (AAS), Secondary Ion Mass Spectrometry (SIMS), and Infrared Reflection Absorption Spectroscopy (IRAS). The oxides grown in reactor in dilute lithium hydroxide solution, specimens cut from Zircaloy, and Zr-2.5Nb alloy pressure tubes removed from CANDU (Canada Deuterium Uranium, Registered Trademark) reactors showed low concentrations of lithium (4 to 50 ppm). The lithium was not leachable in a warm dilute acid. 6 Li undergoes transmutation by the 6 Li(n,t) 4 He reaction. However, SIMS profiles for d 7 Li were identical through the bulk oxide and the isotopic ratio was close to the natural abundance value. The lithium in the oxide, existing as adsorbed lithium on the surface, has been in dynamic equilibrium with lithium in the coolant, and, in spite of many Effective Full Power Years (EFPY) of operation, lithium added to the CANDU coolant at ∼2.5 ppm is not concentrating in the oxides. On the other hand, corrosion films grown in the laboratory in concentrated lithium hydroxide solutions were very porous and contained hundreds of ppm of lithium in the oxide

Development of Self-Healing Zirconium-Silicide Coatings for Improved Performance Zirconium-Alloy Fuel Cladding

Energy Technology Data Exchange (ETDEWEB)

Sridharan, Kumar [University of Wisconsin-Madison; Mariani, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bai, Xianming [Idaho National Lab. (INL), Idaho Falls, ID (United States); Xu, Peng [Westinghouse Electric Company; Lahoda, Ed [Westinghouse Electric Company

2018-03-31

Given the long-term goal of developing such coatings for use with nuclear reactor fuel cladding, this work describes results of oxidation and corrosion behavior of bulk zirconium-silicide and fabrication of zirconium-silicide coatings on zirconium-alloy test flats, tube configurations, and SiC test flats. In addition, boiling heat transfer of these modified surfaces (including ZrSi2 coating) during clad quenching experiments is discussed in detail. Oxidation of bulk ZrSi2 was found to be negligible compared to Zircaloy-4 (a common Zr-alloy cladding material) and mechanical integrity of ZrSi2 was superior to that of bulk Zr2Si at high temperatures in ambient air. Very interesting and unique multi-nanolayered composite of ZrO2 and SiO2 were observed. Physical model for the oxidation has been proposed wherein Zr–Si–O mixture undergoes a spinodal phase decomposition into ZrO2 and SiO2, which is manifested as a nanoscale assembly of alternating layer of the two oxides. Steam corrosion at high pressure (10.3 MPa) led to weight loss of ZrSi2 and produced oxide scale with depletion of silicon, possibly attributed to volatile silicon hydroxide, gaseous silicon monoxide, and a solubility of silicon dioxide in water. Only Zircon phase (ZrSiO4) formed during oxidation of ZrSi2 at 1400°C in air, and allowed for immobilization silicon species in oxide scale in the aqueous environments. Zirconium-silicide coatings (on zirconium-alloy substrates) investigated in this study were deposited primarily using magnetron sputter deposition method and slurry method, although powder spray deposition processes cold spray and thermal spray methods were also investigated. The optimized ZrSi2 sputtered coating exhibited a highly protective nature at elevated temperatures in ambient air by mitigating oxygen permeation to the underlying zirconium alloy substrate. The high oxidation resistance of the coating has been shown to be due to nanocrystalline SiO2 and ZrSiO4 phases in the amorphous

Influence de l’irradiation et de la radiolyse sur la vitesse et les mécanismes de corrosion des alliages de zirconium

OpenAIRE

Verlet , Romain

2015-01-01

The nuclear fuel of pressurized water reactors (PWR) in the form of uranium oxide UO2 pellets (or MOX) is confined in a zirconium alloy cladding. This cladding is very important because it represents the first containment barrier against the release of fission products generated by the nuclear reaction to the external environment. Corrosion by the primary medium of zirconium alloys, particularly the Zircaloy-4, is one of the factors limiting the reactor residence time of the fuel rods (UO2 pe...

Zirconium and hafnium

Science.gov (United States)

Jones, James V.; Piatak, Nadine M.; Bedinger, George M.; Schulz, Klaus J.; DeYoung,, John H.; Seal, Robert R.; Bradley, Dwight C.

2017-12-19

Zirconium and hafnium are corrosion-resistant metals that are widely used in the chemical and nuclear industries. Most zirconium is consumed in the form of the main ore mineral zircon (ZrSiO4, or as zirconium oxide or other zirconium chemicals. Zirconium and hafnium are both refractory lithophile elements that have nearly identical charge, ionic radii, and ionic potentials. As a result, their geochemical behavior is generally similar. Both elements are classified as incompatible because they have physical and crystallochemical properties that exclude them from the crystal lattices of most rock-forming minerals. Zircon and another, less common, ore mineral, baddeleyite (ZrO2), form primarily as accessory minerals in igneous rocks. The presence and abundance of these ore minerals in igneous rocks are largely controlled by the element concentrations in the magma source and by the processes of melt generation and evolution. The world’s largest primary deposits of zirconium and hafnium are associated with alkaline igneous rocks, and, in one locality on the Kola Peninsula of Murmanskaya Oblast, Russia, baddeleyite is recovered as a byproduct of apatite and magnetite mining. Otherwise, there are few primary igneous deposits of zirconium- and hafnium-bearing minerals with economic value at present. The main ore deposits worldwide are heavy-mineral sands produced by the weathering and erosion of preexisting rocks and the concentration of zircon and other economically important heavy minerals, such as ilmenite and rutile (for titanium), chromite (for chromium), and monazite (for rare-earth elements) in sedimentary systems, particularly in coastal environments. In coastal deposits, heavy-mineral enrichment occurs where sediment is repeatedly reworked by wind, waves, currents, and tidal processes. The resulting heavy-mineral-sand deposits, called placers or paleoplacers, preferentially form at relatively low latitudes on passive continental margins and supply 100 percent of

An overview of microstructural and experimental factors that affect the irradiation growth behavior of zirconium alloys

International Nuclear Information System (INIS)

Fidleris, V.; Tucker, R.P.; Adamson, R.B.

1987-01-01

This paper presents an overview of factors affecting irradiation growth of zirconium alloys. Recent data obtained from irradiation programs in EBR-II, ATR, and NRU reactors are used to illustrate the effects of various microstructural and experimental factors on the growth of Zircaloy, zirconium, and zirconium-biobium alloys irradiated to fluences up to 2 X 10 26 nm -2 (E > 1 MeV) over the temperature range 330 to 720 K. Open literature results are also used to confirm or illustrate various effects. Important factors are texture, grain boundary parameters, residual stresses, original dislocation density, microstructure evolution, temperature during irradiation, solute effects, and fluence

Atomic energy

International Nuclear Information System (INIS)

Ramanna, R.

1978-01-01

Development of nuclear science in India, particularly the research and development work at the Bhabha Atomic Research Centre (BARC), Bombay, is described. Among the wide range of materials developed for specific functions under rigorous conditions are nuclear pure grade uranium, zirconium and beryllium, and conventional materials like aluminium, carbon steel and stainless steels. Radioisotopes are produced and used for tracer studies in various fields. Various types of nuclear gauges and nuclear instruments are produced. Radiations have been used to develop new high yielding groundnut mutants with large kernals. The sterile male technique for pest control and radiosterilization technique to process potatoes, onions and marine foods for storage are ready for exploitation. Processes and equipment have been developed for production of electrolytic hydrogen, electrothermal phosphorus and desalinated water. Indigenously manufactured components and materials are now being used for the nuclear energy programme. Indian nuclear power programme strategy is to build heavy water reactors and to utilise their byproduct plutonium and depleted uranium to feed fast breeder reactors which will produce more fissile material than burnt. Finally a special mention has been made of the manpower development programme of the BARC. BARC has established a training school in 1957 giving advanced training in physics, chemistry and various branches of engineering and metallurgy

Structure of zirconium dioxide based porous glasses

Czech Academy of Sciences Publication Activity Database

Gubanova, N. N.; Kopitsa, G. P.; Ezdakova, K. V.; Baranchikov, A. Y.; Angelov, Borislav; Feoktystov, A.; Pipich, V.; Ryukhtin, Vasyl; Ivanov, V. K.

2014-01-01

Roč. 8, č. 5 (2014), s. 967-975 ISSN 1027-4510 R&D Projects: GA ČR GAP208/10/1600; GA MŠk(XE) LM2011019; GA ČR GB14-36566G Institutional support: RVO:61389013 ; RVO:61389005 Keywords : zirconium dioxide * porous glasse * nanoparticles Subject RIV: CF - Physical ; Theoretical Chemistry; BG - Nuclear, Atomic and Molecular Physics, Colliders (UJF-V) Impact factor: 0.359, year: 2012

Accelerators and nuclear reactors as tools in hot atom chemistry

International Nuclear Information System (INIS)

Lindner, L.

1975-01-01

The characteristics of accelerators and of nuclear reactors - the latter to a lesser extent - are discussed in view of their present and future use in hot atom chemistry research and its applications. (author)

Modeling of Some Physical Properties of Zirconium Alloys for Nuclear Applications in Support of UFD Campaign

Energy Technology Data Exchange (ETDEWEB)

Michael V. Glazoff

2013-08-01

Zirconium-based alloys Zircaloy-2 and Zircaloy-4 are widely used in the nuclear industry as cladding materials for light water reactor (LWR) fuels. These materials display a very good combination of properties such as low neutron absorption, creep behavior, stress-corrosion cracking resistance, reduced hydrogen uptake, corrosion and/or oxidation, especially in the case of Zircaloy-4. However, over the last couple of years, in the post-Fukushima Daiichi world, energetic efforts have been undertaken to improve fuel clad oxidation resistance during off-normal temperature excursions. Efforts have also been made to improve upon the already achieved levels of mechanical behavior and reduce hydrogen uptake. In order to facilitate the development of such novel materials, it is very important to achieve not only engineering control, but also a scientific understanding of the underlying material degradation mechanisms, both in working conditions and in storage of used nuclear fuel. This report strives to contribute to these efforts by constructing the thermodynamic models of both alloys; constructing of the respective phase diagrams, and oxidation mechanisms. A special emphasis was placed upon the role of zirconium suboxides in hydrogen uptake reduction and the atomic mechanisms of oxidation. To that end, computational thermodynamics calculations were conducted concurrently with first-principles atomistic modeling.

Method of reducing zirconium

International Nuclear Information System (INIS)

Megy, J.A.

1980-01-01

A method was developed for making nuclear-grade zirconium from a zirconium compound, which ismore economical than previous methods since it uses aluminum as the reductant metal rather than the more expensive magnesium. A fused salt phase containing the zirconium compound to be reduced is first prepared. The fused salt phase is then contacted with a molten metal phase which contains aluminum and zinc. The reduction is effected by mutual displacment. Aluminum is transported from the molten metal phase to the fused salt phase, replacing zirconium in the salt. Zirconium is transported from the fused salt phase to the molten metal phase. The fused salt phase and the molten metal phase are then separated, and the solvent metal and zirconium are separated by distillation or other means. (DN)

Review of experimental studies of zirconium carbide coated fuel particles for high temperature gas-cooled reactors

International Nuclear Information System (INIS)

Minato, Kazuo; Ogawa, Toru; Fukuda, Kousaku

1995-03-01

Experimental studies of zirconium carbide(ZrC) coated fuel particles were reviewed from the viewpoints of fuel particle designs, fabrication, characterization, fuel performance, and fission product retentiveness. ZrC is known as a refractory and chemically stable compound, so ZrC is a candidate to replace the silicon carbide(SiC) coating layer of the Triso-coated fuel particles. The irradiation experiments, the postirradiation heating tests, and the out-of-reactor experiments showed that the ZrC layer was less susceptible than the SiC layer to chemical attack by fission products and fuel kernels, and that the ZrC-coated fuel particles performed better than the standard Triso-coated fuel particles at high temperatures, especially above 1600degC. The ZrC-coated fuel particles demonstrated better cesium retention than the standard Triso-coated fuel particles though the ZrC layer showed a less effective barrier to ruthenium than the SiC layer. (author) 51 refs

High energy beam thermal processing of alpha zirconium alloys and the resulting articles

International Nuclear Information System (INIS)

Sabol, G.P.; McDonald, S.G.; Nurminen, J.I.

1983-01-01

Alpha zirconium alloy fabrication methods and resultant products exhibiting improved high temperature, high pressure steam corrosion resistance. The process, according to one aspect of this invention, utilizes a high energy beam thermal treatment to provide a layer of beta treated microstructure on an alpha zirconium alloy intermediate product. The treated product is then alpha worked to final size. According to another aspect of the invention, high energy beam thermal treatment is used to produce an alpha annealed microstructure in a Zircaloy alloy intermediate size or final size component. The resultant products are suitable for use in pressurized water and boiling water reactors

Effect of alkali chloride and fluoride fused electrolytes on the morphology and size of zirconium deposit

International Nuclear Information System (INIS)

Li, Hui; Park, Kyoung Tae; Kim, Seung Hyun; Lee, Tae Hyuk; Lee, Sang Ki; Lee, Jong Hyeon

2011-01-01

Nuclear-grade, hafnium-free zirconium metal is used in nuclear reactors for structural purposes and for containing the nuclear fuel. The oxygen content is critical as 0.7 %. So the zirconium produced should be in the form of dense, coherent deposits with very low salt content. In this study, the electrorefining of zirconium metal in LiF-KF melts with addition of ZrF 4 has been investigated. During the electrorefining process, the K 2 ZrF 6 was formed and dense and crystalline depositions were obtained. Cyclic voltammetry was used to study the reduction process of Zr ions and the effect of current density to the morphology of depositions was also studied. (author)

ZIRCONIUM-CLADDING OF THORIUM

Science.gov (United States)

Beaver, R.J.

1961-11-21

A method of cladding thorium with zirconium is described. The quality of the bond achieved between thorium and zirconium by hot-rolling is improved by inserting and melting a thorium-zirconium alloy foil between the two materials prior to rolling. (AEC)

Control of microstructure during hot working of zirconium alloys

International Nuclear Information System (INIS)

Chakravartty, J.K.; Banerjee, S.

2005-01-01

Hot working is considered to be the most important step involved in the fabrication of zirconium alloys for nuclear reactor applications for two reasons: i) the scale of the microstructure and texture of the final product is decided at this stage and ii) the hot deformed microstructure provides a suitable starting microstructure for the subsequent fabrication steps. The resultant microstructure in turn controls the properties of the final product. In order to obtain final product with a suitable microstructure and with specified mechanical properties on a repeatable basis the control of microstructure during hot working is of paramount importance. This is usually done by studying the constitutive behaviour of the material under hot working conditions and by constructing processing maps. In the latter method, strain rate sensitivity is mapped as a function of temperature and strain rate to delineate domains within the bounds of which a specific deformation mechanism dominates. Detail microstructural analysis is then carried out on the samples deformed within the domains. Using this methodology, processing maps have been constructed for various zirconium alloys. These maps have been found to be very useful for optimizing the hot workability and control of microstructure of zirconium alloys. (author)

Recovery of zirconium from pickling solution, regeneration and its reuse

Energy Technology Data Exchange (ETDEWEB)

Bhattacharjee, D. [Nuclear Fuel Complex, Hyderabad 500062 (India); Mandal, D., E-mail: dmandal10@gmail.com [Alkali Material & Metal Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Visweswara Rao, R.V.R.L.; Sairam, S.; Thakur, S. [Nuclear Fuel Complex, Hyderabad 500062 (India)

2017-05-15

Graphical abstract: The following compares the performance of fresh pickling solution (PS) and regenerated and used pickling solution (UPS). - Highlights: • Pickling of zircaloy tubes and appendages is carried out to remove oxide layer. • The pickling solution become saturated with zirconium due to reuse. • As NaNO{sub 3} concentration increases, conc. of Zr in pickling solution decreases. • Experimental results shows that, used pickling solution can be regenerated. • Regenerated solution may be reused by adding makeup quantities of HF-HNO{sub 3}. - Abstract: The pressurized heavy water reactors use natural uranium oxide (UO{sub 2}) as fuel and uses cladding material made up of zircaloy, an alloy of zirconium. Pickling of zircaloy tubes and appendages viz., spacer and bearing pads is carried out to remove the oxide layer and surface contaminants, if present. Pickling solution, after use for many cycles i.e., used pickling solution (UPS) is sold out to vendors, basically for its zirconium value. UPS, containing a relatively small concentration of hydrofluoric acid. After repeated use, pickling solution become saturated with zirconium fluoride complex and is treated by adding sodium nitrate to precipitate sodium hexafluro-zirconate. The remaining solution can be recycled after suitable makeup for further pickling use. The revenue lost by selling UPS is very high compared to its zirconium value, which causes monetary loss to the processing unit. Experiments were conducted to regenerate and reuse UPS which will save a good amount of revenue and also protect the environment. Experimental details and results are discussed in this paper.

Development of Zirconium alloys (for pressure tubes)

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Suk; Kwon, Sang Chul; Choo, Ki Nam; Jung, Chung Hwan; Yim, Kyong Soo; Kim, Sung Soo; Baek, Jong Hyuk; Jeong, Yong Hwan; Kim, Kyong Ho; Cho, Hae Dong [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of); Hwang, S. K.; Kim, M. H. [Inha Univ., Incheon (Korea, Republic of); Kwon, S. I [Korea Univ., Seoul (Korea, Republic of); Kim, I. S. [Korea Advanced Inst. of Science and Technology, Taejon (Korea, Republic of)

1997-09-01

The objective of this research is to set up the basic technologies for the evaluation of pressure tube integrity and to develop improved zirconium alloys to prevent pressure tube failures due to DHC and hydride blister caused by excessive creep-down of pressure tubes. The experimental procedure and facilities for characterization of pressure tubes were developed. The basic research related to a better understanding of the in-reactor performances of pressure tubes leads to noticeable findings for the first time : the microstructural effect on corrosion and hydrogen pick-up behavior of Zr-2.5Nb pressure tubes, texture effect on strength and DHC resistance and enhanced recrystallization by Fe in zirconium alloys and etc. Analytical methodology for the assessment of pressure tubes with surface flaws was set up. A joint research is being under way with AECL to determine the fracture toughness of O-8 at the EOL (End of Life) that had been quadruple melted and was taken out of the Wolsung Unit-1 after 10 year operation. In addition, pressure tube with texture controlled is being made along with VNINM in Russia as a joint project between KAERI and Russia. Finally, we succeeded in developing 4 different kinds of zirconium alloys with better corrosion resistance, low hydrogen pickup fraction and higher creep strength. (author). 121 refs., 65 tabs., 260 figs

Management of waste cladding hulls. Part II. An assessment of zirconium pyrophoricity and recommendations for handling waste hulls

International Nuclear Information System (INIS)

Kullen, B.J.; Levitz, N.M.; Steindler, M.J.

1977-11-01

This report reviews experience and research related to the pyrophoricity of zirconium and zirconium alloys. The results of recent investigations of the behavior of Zircaloy and some observations of industrial handling and treatment of Zircaloy tubing and scrap are also discussed. A model for the management of waste Zircaloy cladding hulls from light water reactor fuel reprocessing is offered, based on an evaluation of the reviewed information. It is concluded that waste Zircaloy cladding hulls do not constitute a pyrophoric hazard if, following the model flow sheet, finely divided metal is oxidized during the management procedure. Steps alternative to the model are described which yield zirconium in deactivated form and also accomplish varying degrees of transuranic decontamination. Information collected into appendixes is (1) a collation of zirconium pyrophoricity data from the literature, (2) calculated radioactivity contents in Zircaloy cladding hulls from spent LWR fuels, and (3) results of a laboratory study on volatilization of zirconium from Zircaloy using HCl or Cl 2

Influence of irradiation and radiolysis on the corrosion rates and mechanisms of zirconium alloys

International Nuclear Information System (INIS)

Verlet, Romain

2015-01-01

The nuclear fuel of pressurized water reactors (PWR) in the form of uranium oxide UO 2 pellets (or MOX) is confined in a zirconium alloy cladding. This cladding is very important because it represents the first containment barrier against the release of fission products generated by the nuclear reaction to the external environment. Corrosion by the primary medium of zirconium alloys, particularly the Zircaloy-4, is one of the factors limiting the reactor residence time of the fuel rods (UO 2 pellets + cladding). To optimize core management and to extend the lifetime of the fuel rods in reactor, new alloys based on zirconium-niobium (M5) have been developed. However, the corrosion mechanisms of these are not completely understood because of the complexity of these materials, corrosion environment and the presence of radiation from the nuclear fuel. Therefore, this thesis specifically addresses the effects of radiolysis and defects induced by irradiation with ions in the matrix metal and the oxide layer on the corrosion rate of Zircaloy-4 and M5. The goal is to separate the influence of radiation damage to the metal, that relating to defects created in the oxide and that linked to radiolysis of the primary medium on the oxidation rate of zirconium alloys in reactor. 1) Regarding effect of irradiation of the metal on the oxidation rate: type dislocation loops appear and increase the oxidation rate of the two alloys. For M5, in addition to the first effect, a precipitation of fines needles of niobium reduced the solid solution of niobium concentration in the metal and ultimately in the oxide, which strongly reduces the oxidation rate of the alloy. 2) Regarding the effect of irradiation of the oxide layer on the oxidation rate: defects generated by the nuclear cascades in the oxide increase the oxidation rate of the two materials. For M5, germination of niobium enriched zones in irradiated oxide also causes a decrease of the niobium concentration in solid solution

Metallurgy of zirconium and hafnium

International Nuclear Information System (INIS)

Baryshnikov, N.V.; Geger, V.Eh.; Denisova, N.D.; Kazajn, A.A.; Kozhemyakin, V.A.; Nekhamkin, L.G.; Rodyakin, V.V.; Tsylov, Yu.A.

1979-01-01

Considered are those properties of zirconium and of hafnium, which are of practical interest for the manufacture of these elements. Systematized are the theoretical and the practical data on the procedures for thermal decomposition of zirconia and for obtaining zirconium dioxide and hafnium dioxide by a thermal decomposition of compounds and on the hydrometallurgical methods for extracting zirconium and hafnium. Zirconium and hafnium fluorides and chlorides production procedures are described. Considered are the iodide and the electrolytic methods of refining zirconium and hafnium

On elastic structural elements for nuclear reactors

International Nuclear Information System (INIS)

Povolo, F.

1978-03-01

The in-pile stress-relaxation behaviour of materials usually employed for the elastic structural elements, in nuclear reactors, is critically reviewed and the results are compared with those obtained in commercial zirconium alloys irradiated under similar conditions. Finally, it is shown that, under certain conditions, some zirconium alloys may be used as an alternative material for these structural elements. (orig.) [de

The behaviour of zirconium alloys in Santowax OM organic coolant at high temperatures

International Nuclear Information System (INIS)

Sawatzky, A.

1964-10-01

Zirconium alloys have been exposed to Santowax OM at temperatures of 320 to 400 o C for times as long as 5000 hours. Short-term experiments (less than 2 weeks) were done in stainless-steel bombs and small out-of-pile loops. The X-7 organic loop in the NRX reactor was used to study long-term oxidation and hydriding both in-flux and out-of-flux. The results obtained lead to several tentative conclusions: Aluminum cladding serves as an effective hydrogen barrier; Considerable protection against hydriding is given by zirconium oxide, provided impurities in the organic are carefully controlled; Hydriding is greatly enhanced by the presence of chlorine in the coolant; and, Hydriding is somewhat enhanced by neutron irradiation. Of considerable significance is the fact that a Zircaloy-4 in-reactor test section of the X-7 loop was exposed to Santowax OM at 320 to 400 o C for more than 5000 hours without excessive hydriding. (author)

Synthesis of zirconium by zirconium tetrachloride reduction by magnesio-thermia. Experimental study and modelling; Elaboration de zirconium par reduction de tetrachlorure de zirconium par magnesothermie. Etude experimentale et modelisation

Energy Technology Data Exchange (ETDEWEB)

Basin, N

2001-01-01

This work deals with the synthesis of zirconium. The ore is carbo-chlorinated to obtain the tetrachloride which is then purified by selective condensation and extractive distillation. Zirconium tetrachloride is then reduced by magnesium and the pseudo-alloy is obtained according to the global following reaction (Kroll process): ZrCl{sub 4} + 2 Mg = 2 MgCl{sub 2}. By thermodynamics, it has been shown that the volatilization of magnesium chloride and the formation of zirconium sub-chlorides are minimized when the combined effects of temperature and of dilution with argon are limited. With these conditions, the products, essentially zirconium and magnesium chloride, are obtained in equivalence ratio in the magnesio-thermia reaction. The global kinetics of the reduction process has been studied by a thermal gravimetric method. A thermo-balance device has been developed specially for this kinetics study. It runs under a controlled atmosphere and is coupled to a vapor tetrachloride feed unit. The transformation is modelled supposing that the zirconium and magnesium chloride formation result: 1)of the evaporation of magnesium from its liquid phase 2)of the transfer of magnesium and zirconium tetrachloride vapors towards the front of the reaction located in the gaseous phase 3)of the chemical reaction. In the studied conditions, the diffusion is supposed to be the limiting process. The influence of the following parameters: geometry of the reactive zone, temperature, scanning rate of the argon-zirconium tetrachloride mixture, composition of the argon-zirconium tetrachloride mixture has been experimentally studied and confronted with success to the model. (O.M.)

Process for purifying zirconium sponge

International Nuclear Information System (INIS)

Abodishish, H.A.M.; Kimball, L.S.

1992-01-01

This patent describes a Kroll reduction process wherein a zirconium sponge contaminated with unreacted magnesium and by-product magnesium chloride is produced as a regulus, a process for purifying the zirconium sponge. It comprises: distilling magnesium and magnesium chloride from: a regulus containing a zirconium sponge and magnesium and magnesium chloride at a temperature above about 800 degrees C and at an absolute pressure less than about 10 mmHg in a distillation vessel to purify the zirconium sponge; condensing the magnesium and the magnesium chloride distilled from the zirconium sponge in a condenser; and then backfilling the vessel containing the zirconium sponge and the condenser containing the magnesium and the magnesium chloride with a gas; recirculating the gas between the vessel and the condenser to cool the zirconium sponge from above about 800 degrees C to below about 300 degrees C; and cooling the recirculating gas in the condenser containing the condensed magnesium and the condensed magnesium chloride as the gas cools the zirconium sponge to below about 300 degrees C

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Hindle, E.D.

1984-01-01

The fuel elements for a pressurised water reactor comprise arrays of rods of zirconium alloy sheathed nuclear fuel pellets. The helium gas pressure within each rod differs substantially from that of its closest neighbours

Irradiation creep in zirconium single crystals

International Nuclear Information System (INIS)

MacEwen, S.R.; Fidleris, V.

1976-07-01

Two identical single crystals of crystal bar zirconium have been creep tested in reactor. Both specimens were preirradiated at low stress to a dose of about 4 x 10 23 n/m 2 (E > 1 MeV), and were then loaded to 25 MPa. The first specimen was loaded with reactor at full power, the second during a shutdown. The loading strain for both crystals was more than an order of magnitude smaller than that observed when an identical unirradiated crystal was loaded to the same stress. Both crystals exhibited periods of primary creep, after which their creep rates reached nearly constant values when the reactor was at power. During shutdowns the creep rates decreased rapidly with time. Electron microscopy revealed that the irradiation damage consisted of prismatic dislocation loops, approximately 13.5 nm in diameter. Cleared channels, identified as lying on (1010) planes, were also observed. The results are discussed in terms of the current theories for flux enhanced creep in the light of the microstructures observed. (author)

Recent irradiation tests of uranium-plutonium-zirconium metal fuel elements

International Nuclear Information System (INIS)

Pahl, R.G.; Lahm, C.E.; Villarreal, R.; Hofman, G.L.; Beck, W.N.

1986-09-01

Uranium-Plutonium-Zirconium metal fuel irradiation tests to support the ANL Integral Fast Reactor concept are discussed. Satisfactory performance has been demonstrated to 2.9 at.% peak burnup in three alloys having 0, 8, and 19 wt % plutonium. Fuel swelling measurements at low burnup in alloys to 26 wt % plutonium show that fuel deformation is primarily radial in direction. Increasing the plutonium content in the fuel diminishes the rate of fuel-cladding gap closure and axial fuel column growth. Chemical redistribution occurs by 2.1 at.% peak burnup and generally involves the inward migration of zirconium and outward migration of uranium. Fission gas release to the plenum ranges from 46% to 56% in the alloys irradiated to 2.9 at.% peak burnup. No evidence of deleterious fuel-cladding chemical or mechanical interaction was observed

Collaborative analysis for certification of zirconium and zirconium base alloy reference materials JAERI-Z11 to Z16

International Nuclear Information System (INIS)

1985-03-01

The second Sub-Committee on Zircaloy Analysis was organized in April 1978, under the Committee on Analytical Chemistry on Nuclear Fuels and Reactor Materials, JAERI, for the renewal of zirconium and zirconium base alloy certified reference materials (CRMs). The Sub-Committee carried out collaborative analysis among 13 participating laboratories for the certification of the CRMs, JAERI-Z11 to Z18, after development, improvement and evaluation of analytical methods during the period of May 1978 to June 1982. As the result of the collaborative analysis, the certified value was given for 18 elements (Sn, Fe, Ni, Cr, B, Cd, U, Cu, Co, Mn, Pb, Al, Ti, Si, Mo, W, Hf, C) in the CRMs. The first part of this report includes general discussion, the second part principles of certification, the third part development and verification of analytical methods, and the fourth part evaluation of analytical results on 17 elements. Preparation of Z11 to Z18, and certification for carbon in JAERI-Z17 and Z18 were reported separately in JAERI-M 83-241 and M 83-035, respectively. (author)

Diffusion of insoluble carbon in zirconium oxides

CERN Document Server

Vykhodets, V B; Koester, U; Kondrat'ev, V V; Kesarev, A G; Hulsen, C; Kurennykh, T E

2011-01-01

The diffusion coefficient of insoluble carbon in zirconium oxides has been obtained for the temperature range of 900-1000A degrees C. There are no published data on the diffusion of insoluble impurities; these data are of current interest for the diffusion theory and nuclear technologies. Tracer atoms 13C have been introduced into oxides by means of ion implantation and the kinetics of their emission from the samples in the process of annealing in air has been analyzed. The measurements have been performed using the methods of nuclear microanalysis and X-ray photoelectron spectroscopy. The diffusion activation energy is 2.7 eV and the carbon diffusion coefficient is about six orders of magnitude smaller than that for oxygen self-diffusion in the same systems. This result indicates the strong anomaly of the diffusion properties of carbon in oxides. As a result, zirconium oxides cannot be used in some nuclear technologies, in particular, as a material of sources for accelerators of short-lived carbon isotopes.

Contribution to the study of zirconium self-diffusion in zirconium carbide

International Nuclear Information System (INIS)

An, Chul

1972-01-01

The objective of this research thesis is to determine experimental conditions allowing the measurement of the self-diffusion coefficient of zirconium in zirconium carbide. The author reports the development of a method of preparation of zirconium carbide samples. He reports the use of ion implantation as technique to obtain a radio-tracer coating. The obtained results give evidence of the impossibility to use sintered samples with small grains because of the demonstrated importance of intergranular diffusion. The self-diffusion coefficient is obtained in the case of zirconium carbide with grains having a diameter of few millimetres. The presence of 95 Nb from the disintegration of 95 Zr indicates that these both metallic elements have very close diffusion coefficients at 2.600 C [fr

Preliminary radiation-oxidizing treatment influence on radiation-catalytic activity of zirconium during water decomposition process

International Nuclear Information System (INIS)

Garibov, A.A.; Aliyev, A.G.; Agayev, T.N.; Aliyev, S.M.; Velibekova, G.Z.

2004-01-01

The study of physical-chemical processes proceeding in contact of metal constructional materials nuclear reactors with water at simultaneous influence of temperature and radiation represents the large interest at the decision of problems material authority and safety of work of nuclear -power installations [1-2]. One of the widely widespread materials of active zone nuclear reactors is metal zirconium and its alloys. The influence of preliminary radiation processing on radiation, radiation -thermal and thermal processes of accumulation of molecular hydrogen and oxidation zirconium in contact with water is investigated at T=673 K and ρ=5mg/sm 3 [3-4]. Initial samples zirconium previously has been exposed by an irradiation in medium H 2 O 2 at D=20-410 kGy. The contribution of radiation processes in these contacts in process thermo-radiation decomposition of water and oxidation of materials of zirconium is revealed. It is established that the interaction of Zr metal, preliminary treated by radiation, with water at radiation -heterogeneous processes leads to passivity of a surface. The rate meanings of thermal, radiation -thermal processes and radiation-chemical yields of hydrogen are determined. It is revealed, that at radiation-heterogeneous processes in system Zr +H 2 O (ρ =5mg/sm 3 T=673 K) the increase of the absorbed doze up to 123 kGy results to reduction of a radiation -chemical yield of molecular hydrogen. The further increase of the absorbed doze results to increase of a radiation -chemical yield of hydrogen. The observable effect at the preliminary radiation of zirconium is connected to formation of oxide phase on a surface. The mechanism of radiation -heterogeneous processes proceeding in system Zr+H 2 O is suggested. (author)

Environmentally-induced cracking of zirconium alloys - a review

International Nuclear Information System (INIS)

Cox, B.

1990-01-01

The general field of environmentally-induced cracking of zirconium alloys has been reviewed and the phenomena that are observed and the progress in understanding the mechanisms are summarized. The details of the industrially important pellet-clad interaction failures of nuclear reactor fuel have been left for a companion review, and only observations on the mechanism are summarized briefly here. It is concluded that in the zirconium alloy system, by virtue of the physical peculiarities of the system, it is easier to reach unambiguous conclusions about the environmental cracking mechanisms that are operating than with other systems. Thus, chemical dissolution in either liquid or vapour phase is thought to be the principal mechanism for intergranular cracking, while adsorption-induced embrittlement is thought to be the most common transgranular quasi-cleavage process. Hydrogen embrittlement in this system can be identified because it requires precipitated hydride that gives characteristic fractography when cracked. Only in a few instances does stress-corrosion cracking appear to proceed by a hydride cracking mechanism. (orig.)

Atomization of U3Si2 for research reactor fuel

International Nuclear Information System (INIS)

Kim, C.K.; Kim, K.H.; Lee, C.T.; Kuk, I.H.

1995-01-01

Rotating disk atomization technique is applied to KMRR (Korea Multi-purpose Research Reactor) fuel fabrication. A rotating disk atomizer is designed and manufactured locally and U-4.0 wt. % Si alloy powders are produced. The atomized powders are heat-treated to transform into U 3 Si and the mixture of U 3 Si and Al are extruded to fuel meat. Most of the atomized powders are spherical in shape. The microstructure of the powder is fine due to the rapid solidification. The time required for peritectoid reaction is reduced due to the fine microstructures and the resultant U 3 Si grain size is finer than ever obtained from ingot process. The mechanical properties of the fuel meat are improved: yield strength about 30 %, tensile strength 10% and elongation 250 % increased. (author)

Polycrystalline models for the calculation of residual stresses in zirconium alloys tubes

International Nuclear Information System (INIS)

Signorelli, J.W.; Turner, P.A.; Lebensohn, R.A.; Pochettino, A.A.

1995-01-01

Tubes made of different Zirconium alloys are used in various types of reactors. The final texture of tubes as well as the distribution of residual stresses depend on the mechanical treatments done during their manufacturing process. The knowledge and prediction of both the final texture and the distribution of residual stresses in a tube for nuclear applications are of outstanding importance in relation with in-reactor performance of the tube, especially in what concerns to its irradiation creep and growth behaviour. The viscoplastic and the elastoplastic self consistent polycrystal models are used to investigate the influence of different mechanical treatments, performed during rolling processes on the final distribution of intergranular residual stresses of zirconium alloys tubes. The residual strains predictions with both formulations show a non linear dependence with the orientation, but they are qualitatively different. This discrepancy could be explain in terms of the relative plastic activity between the -type and -type deformation modes predicted with the viscoplastic and elastoplastic models. (author). 10 refs., 4 figs., 1 tab

Estimation of niobium, tantalum and zirconium in a carbonate rich ore

International Nuclear Information System (INIS)

Navale, A.S.; Venkatakrishnan, R.R.; Sreenivas, T.

2013-01-01

Tantalum, niobium and zirconium form refractory compounds and therefore pose problems in estimation by methods like spectrophotometry, AAS etc. Their estimation is best carried out with ICP-OES as the plasma is capable of breaking the refractory compounds and releases these elements in the form of excited atoms. But another major problem is bringing these elements in solution as they require very strong reagents and tend to hydrolyze easily in solution. Complexing agents need to be added to maintain these elements in dissolved condition. The commonly employed dissolution techniques for these elements are fusion with potassium pyrosulfate (for Nb and Ta), sodium hydroxide or peroxide (for zirconium). Digestion with mineral acids in combination with hydrofluoric acid is also employed

Materials for nuclear reactors

International Nuclear Information System (INIS)

Banerjee, S.; Kamath, H.S.

2005-01-01

The improved performance of present generation nuclear reactors and the realization of advanced reactor concepts, both, require development of better materials. Physical metallurgy/materials science principles which have been exploited in meeting the exacting requirements of nuclear reactor materials (fuels and structural materials), are outlined citing a few specific examples. While the incentive for improvement of traditional fuels (e.g., UO 2 fuel) is primarily for increasing the average core burn up, the development of advanced fuels (e.g., MOX, mixed carbide, nitride, silicide and dispersion fuels) are directed towards better utilization of fissile and fertile inventories through adaptation of innovative fuel cycles. As the burn up of UO 2 fuel reaches higher levels, a more detailed and quantitative understanding of the phenomena such as fission gas release, fuel restructuring induced by radiation and thermal gradients and pellet-clad interaction is being achieved. Development of zirconium based alloys for both cladding and pressure tube applications is discussed with reference to their physical metallurgy, fabrication techniques and in-reactor degradation mechanisms. The issue of radiation embrittlement of reactor pressure vessels (RPVs) is covered drawing a comparison between the western and eastern specifications of RPV steels. The search for new materials which can stand higher rates of atomic displacement due to radiation has led to the development of swelling resistant austenitic and ferritic stainless steels for fast reactor applications as exemplified by the development of the D-9 steel for Indian fast breeder reactor. The presentation will conclude by listing various materials related phenomena, which have a strong bearing on the successful development of future nuclear energy systems. (author)

Modification of zirconium diphosphate with salicylic acid and its effect on the uranium (Vi) sorption

International Nuclear Information System (INIS)

Almazan T, M. G.; Garcia G, N.; Simoni, E.

2014-10-01

The surface of zirconium diphosphate (ZrP 2 O 7 ) was modified with salicylic acid and its effect was evaluated on the uranium (Vi) sorption. The modified surface of the material was analyzed with different analytical techniques among which are included the atomic force microscopy, scanning electron microscopy and X-ray photoelectron spectroscopy. This analysis allowed showing that the salicylic acid is being held on the surface of the zirconium diphosphate. The reactivity of modified zirconium diphosphate compared with uranium (Vi) was investigated using the classical method of batch sorption. The analysis of sorption isotherms shows that the salicylic acid has an important effect in the uranium (Vi) sorption. According to the study conducted, the interaction among the uranium (Vi) and the surface of zirconium diphosphate modified with the salicylic acid most likely leads to the complexes formation of binary (U(Vi)/ZrP 2 O 7 ) and ternary (U(Vi)/salicylate/ZrP 2 O 7 ) surface. (Author)

Artefacts in multimodal imaging of titanium, zirconium and binary titanium-zirconium alloy dental implants: an in vitro study.

Science.gov (United States)

Smeets, Ralf; Schöllchen, Maximilian; Gauer, Tobias; Aarabi, Ghazal; Assaf, Alexandre T; Rendenbach, Carsten; Beck-Broichsitter, Benedicta; Semmusch, Jan; Sedlacik, Jan; Heiland, Max; Fiehler, Jens; Siemonsen, Susanne

2017-02-01

To analyze and evaluate imaging artefacts induced by zirconium, titanium and titanium-zirconium alloy dental implants. Zirconium, titanium and titanium-zirconium alloy implants were embedded in gelatin and MRI, CT and CBCT were performed. Standard protocols were used for each modality. For MRI, line-distance profiles were plotted to quantify the accuracy of size determination. For CT and CBCT, six shells surrounding the implant were defined every 0.5 cm from the implant surface and histogram parameters were determined for each shell. While titanium and titanium-zirconium alloy induced extensive signal voids in MRI owing to strong susceptibility, zirconium implants were clearly definable with only minor distortion artefacts. For titanium and titanium-zirconium alloy, the MR signal was attenuated up to 14.1 mm from the implant. In CT, titanium and titanium-zirconium alloy resulted in less streak artefacts in comparison with zirconium. In CBCT, titanium-zirconium alloy induced more severe artefacts than zirconium and titanium. MRI allows for an excellent image contrast and limited artefacts in patients with zirconium implants. CT and CBCT examinations are less affected by artefacts from titanium and titanium-zirconium alloy implants compared with MRI. The knowledge about differences of artefacts through different implant materials and image modalities might help support clinical decisions for the choice of implant material or imaging device in the clinical setting.

Artefacts in multimodal imaging of titanium, zirconium and binary titanium–zirconium alloy dental implants: an in vitro study

Science.gov (United States)

Schöllchen, Maximilian; Aarabi, Ghazal; Assaf, Alexandre T; Rendenbach, Carsten; Beck-Broichsitter, Benedicta; Semmusch, Jan; Sedlacik, Jan; Heiland, Max; Fiehler, Jens; Siemonsen, Susanne

2017-01-01

Objectives: To analyze and evaluate imaging artefacts induced by zirconium, titanium and titanium–zirconium alloy dental implants. Methods: Zirconium, titanium and titanium–zirconium alloy implants were embedded in gelatin and MRI, CT and CBCT were performed. Standard protocols were used for each modality. For MRI, line–distance profiles were plotted to quantify the accuracy of size determination. For CT and CBCT, six shells surrounding the implant were defined every 0.5 cm from the implant surface and histogram parameters were determined for each shell. Results: While titanium and titanium–zirconium alloy induced extensive signal voids in MRI owing to strong susceptibility, zirconium implants were clearly definable with only minor distortion artefacts. For titanium and titanium–zirconium alloy, the MR signal was attenuated up to 14.1 mm from the implant. In CT, titanium and titanium–zirconium alloy resulted in less streak artefacts in comparison with zirconium. In CBCT, titanium–zirconium alloy induced more severe artefacts than zirconium and titanium. Conclusions: MRI allows for an excellent image contrast and limited artefacts in patients with zirconium implants. CT and CBCT examinations are less affected by artefacts from titanium and titanium–zirconium alloy implants compared with MRI. The knowledge about differences of artefacts through different implant materials and image modalities might help support clinical decisions for the choice of implant material or imaging device in the clinical setting. PMID:27910719

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

Method of separating hafnium from zirconium

International Nuclear Information System (INIS)

Megy, J.A.

1980-01-01

English. A new anhydrous method was developed for separating zirconium and hafnium, which gives higher separation factors and is more economical than previous methods. A molten phase, comprising a solution of unseparated zirconium and hafnium and a solvent metal, is first prepared. The molten metal phase is contacted with a fused salt phase which includes a zirconium salt. Zirconium and hafnium separation is effected by mutual displacement with hafnium being transported from the molten metal phase to the fused salt phase, while zirconium is transported from the fused salt phase to the molten metal phase. The solvent metal is less electropositive than zirconium. Zinc was chosen as the solvent metal, from a group which also included cadmium, lead, bismuth, copper, and tin. The fused salt phase cations are more electropositive than zirconium and were selected from a group comprising the alkali elements, the alkaline earth elements, the rare earth elements, and aluminum. A portion of the zirconium in the molten metal phase was oxidized by injecting an oxidizing agent, chlorine, to form zirconium tetrachlorid

Development of microstructure in thermomechanical processing of zirconium alloys

International Nuclear Information System (INIS)

Jha, S.K.; Saibaba, N.; Jayaraj, R.N.

2009-01-01

Zirconium based alloys are used for the manufacture of fuel tubes pressure tubes calandria tubes and other components of Pressurized Heavy Water Reactors (PHWRS). In single or two phase zirconium alloy system a variety of microstructure can be generated by suitable heat treatments by the process of equilibrium and non equilibrium phase transformations Microstructure can also be modified by alloying with α and β stabilizers. The microstructure in Zr alloys could be single hexagonal phase (α alloys) two phase bcc and hexagonal (α + β alloys) phase, single metastable martensitic microstructure and β with ω phase. The microstructural and micro textural evolution during thermo mechanical treatments depends strongly on such initial microstructure. Hot extrusion is a significant bulk deformation step which decides the initial microstructure of the alloy. It is carried out at elevated temperature i e above the recrystallization temperature, which enable imposition of large strains in single step. This deformation causes a significant change in the microstructure of the material and depends on extrusion process parameters such as temperature, strain rate (Ram speed), reduction ratio etc. In the present paper development of microstructures, microtexture and texture have been examined. An attempt is also made to optimise the hot working parameters for different Zirconium alloys with help of these studies. (author)

Zirconium-barrier cladding attributes

International Nuclear Information System (INIS)

Rosenbaum, H.S.; Rand, R.A.; Tucker, R.P.; Cheng, B.; Adamson, R.B.; Davies, J.H.; Armijo, J.S.; Wisner, S.B.

1987-01-01

This metallurgical study of Zr-barrier fuel cladding evaluates the importance of three salient attributes: (1) metallurgical bond between the zirconium liner and the Zircaloy substrate, (2) liner thickness (roughly 10% of the total cladding wall), and (3) softness (purity). The effect that each of these attributes has on the pellet-cladding interaction (PCI) resistance of the Zr-barrier fuel was studied by a combination of analytical model calculations and laboratory experiments using an expanding mandrel technique. Each of the attributes is shown to contribute to PCI resistance. The effect of the zirconium liner on fuel behavior during off-normal events in which steam comes in contact with the zirconium surface was studied experimentally. Simulations of loss-of-coolant accident (LOCA) showed that the behavior of Zr-barrier cladding is virtually indistinguishable from that of conventional Zircaloy cladding. If steam contacts the zirconium liner surface through a cladding perforation and the fuel rod is operated under normal power conditions, the zirconium liner is oxidized more rapidly than is Zircaloy, but the oxidation rate returns to the rate of Zircaloy oxidation when the oxide phase reaches the zirconium-Zircaloy metallurgical bond

Experimental studies of relevance on zirconium nitrate raffinate sludge for its disposal as well as zirconium recovery

International Nuclear Information System (INIS)

Brahmananda Reddy, G.; Narasimha Murty, B.; Ravindra, H.R.

2013-01-01

One of the many routes of production of nuclear grade zirconium dioxide involve separation of zirconium and hafnium by solvent extraction of zirconium nitrate using tri-n-butyl phosphate followed by precipitation of zirconium with ammonia and finally calcination of the so obtained hydrated zirconia at elevated temperature. The zirconium feed solution as is generated from digestion of zirconium washed dried frit (produced by the caustic fusion of zircon sand which is one of the beach sand heavy minerals) in nitric acid contain considerable amount of sludge material and after solvent extraction this whole sludge material rests with raffinate. This sludge material has a scope to contain considerable amounts of zirconium along with other metal ions such as hafnium, aluminium, iron, etc. besides nitric acid and it constitutes one of the important solid wastes that needs to be disposed suitably. One of the disposal means of this sludge material is to use it as a land fill for which two important criteria are to be viz the pH of 10% solid waste solution should be near to neutral pH and the loss on ignition at 550℃ on dry basis of the sludge to be below 20%. In order to study the implications of presence of varying amounts of zirconium nitrate in the sludge on the pH of 10% solution of the sludge various synthetic zirconium nitrate solid waste were prepared using the sludge material generated at the laboratory during the analysis of zirconium washed dried frit. Presence of zirconium in the sludge is expected to decrease the overall pH of the 10% solution of the sludge because zirconium is prone to hydrolyze especially locally when zirconium ion comes into contact with water according to the chemical equation Zr 4+ H 2 O → ZrO 2+ + 2H + . From this equation, it is clear that for every one mole of zirconium ions two moles of hydrogen ions are produced. This is verified experimentally using the synthetically prepared sludge materials with varying amounts of zirconium

Small liquid metal reactor for an initial phase of fast breeder reactor introduction

International Nuclear Information System (INIS)

Ishiguro, Y.; Nascimento, J.A. do.

1985-01-01

Safety and burnup characteristics of a 1000 MWth liquid metal reactor have been examined for various fuel types. With metallic Pu/Th fuel containing a small amount of zirconium hydride, low sodium-void reactivity, a high Doppler coefficient, and small burnup reactivity swings can be achieved. A conservative design is considered for an initial phase of fast breeder reactor development and possible modifications are discussed. (Author) [pt

Localized deformation of zirconium-liner tube

International Nuclear Information System (INIS)

Nagase, Fumihisa; Uchida, Masaaki

1988-03-01

Zirconium-liner tube has come to be used in BWR. Zirconium liner mitigates the localized stress produced by the pellet-cladding interaction (PCI). In this study, simulating the ridging, stresses were applied to the inner surfaces of zirconium-liner tubes and Zircaloy-2 tubes, and, to investigate the mechanism and the extent of the effect, the behavior of zirconium liner was examined. As the result of examination, stress was concentrated especially at the edge of the deformed region, where zirconium liner was highly deformed. Even after high stress was applied, the deformation of Zircaloy part was small, since almost the concentrated stress was mitigated by the deformation of zirconium liner. In addition, stress and strain distributions in the cross section of specimen were calculated with a computer code FEMAXI-III. The results also showed that zirconium liner mitigated the localized stress in Zircaloy, although the affected zone was restricted to the region near the boundary between zirconium liner and Zircaloy. (author)

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Hindle, E.D.

1981-01-01

An array of rods comprising zirconium alloy sheathed nuclear fuel pellets assembled to form a fuel element for a pressurised water reactor is claimed. The helium gas pressure within each rod differs substantially from that of its closest neighbours

Zirconium isotope separation process

International Nuclear Information System (INIS)

Peterson, S.H.; Lahoda, E.J.

1988-01-01

A process is described for reducing the amount of zirconium 91 isotope in zirconium comprising: forming a first solution of (a) a first solvent, (b) a scavenger, and (c) a zirconium compound which is soluble in the first solvent and reacts with the scavenger when exposed to light of a wavelength of 220 to 600 nm; irradiating the first solution with light at the wavelength for a time sufficient to photoreact a disproportionate amount of the zirconium compound containing the zirconium 91 isotope with the scavenger to form a reaction product in the first solution; contacting the first solution, while effecting the irradiation, with a second solvent which is immiscible with the first solvent, which the second solvent is a preferential solvent for the reaction product relative to the first solvent, such that at least a portion of the reaction product is transferred to the second solvent to form a second solution; and separating the second solution from the first solution after the contacting

The {sup 99m}Tc generator using poly zirconium chloride (PZC)

Energy Technology Data Exchange (ETDEWEB)

Wan Awang, Wan Anuar; Dahalan, Rehir; Kasim, Shaaban [Medical Technology Div., Malaysian Inst. for Nuclear Technology Research (MINT), Selangor (Malaysia)

1998-10-01

The {sup 99m}Tc generator system has been prepared using poly zirconium chloride (PZC) to replace alumina (Al{sub 2}O{sub 3}) as a adsorbent. The Mo-98 (MoO{sub 3}) was irradiated using our 1 MW TRIGA MARK II research reactor. The amount of Mo-99 adsorbed to the PZC has been studied and the yield from the elution was about 20%. (author)

Unirradiated UO2 in irradiated zirconium alloy sheathing

International Nuclear Information System (INIS)

MacDonald, R.D.; Hardy, D.G.; Hunt, C.E.L.; Scoberg, J.A.

1979-07-01

Zircaloy-clad UO 2 fuel elements have defected in power reactors when element power outputs were raised significantly after a long irradiation at low power. We have irradiated fuel elements fabricated from fresh UO 2 pellets and zirconium alloy sheaths previously irradiated without fuel. This gave a fuel element with radiation-damaged low-ductility sheathing but with no fission products in the fuel. The elements were power boosted in-reactor to linear power outputs up to 84 kW/m for two five-day periods. No elements defected despite sheath strains of 0.82 percent at circumferential ridge postions. Half of these elements were subsequently soaked at low power to build up the fission product inventory in the fuel and then power boosted to 63 kW/m for a third time. Two elements defected on this final boost. We conclude that these defects were caused by fission product induced stress-corrosion cracking and that this mechanism plays an importent role in power reactor fuel defects. (auth)

Zirconium for nitric acid solutions

International Nuclear Information System (INIS)

Yau, T.L.

1984-01-01

The excellent corrosion resistance of zirconium in nitric acid has been known for over 30 years. Recently, there is an increasing interest in using zirconium for nitric acid services. Therefore, an extensive research effort has been carried out to achieve a better understanding of the corrosion properties of zirconium in nitric acid. Particular attention is paid to the effect of concentration, temperature, structure, solution impurities, and stress. Immersion, autoclave, U-bend, and constant strain-rate tests were used in this study. Results of this study indicate that the corrosion resistance of zirconium in nitric acid is little affected by changes in temperature and concentration, and the presence of common impurities such as seawater, sodium chloride, ferric chloride, iron, and stainless steel. Moreover, the presence of seawater, sodium chloride, ferric chloride, and stainless steel has little effect on the stress corrosion craking (SCC) susceptibility of zirconium in 70% nitric acid at room temperatures. However, zirconium could be attacked by fluoride-containing nitric acid and the vapors of chloride-containing nitric acid. Also, high sustained tensile stresses should be avoided when zirconium is used to handle 70% nitric acid at elevated temperatures or > 70% nitric acid

Process for etching zirconium metallic objects

International Nuclear Information System (INIS)

Panson, A.J.

1988-01-01

In a process for etching of zirconium metallic articles formed from zirconium or a zirconium alloy, wherein the zirconium metallic article is contacted with an aqueous hydrofluoric acid-nitric acid etching bath having an initial ratio of hydrofluoric acid to nitric acid and an initial concentration of hydrofluoric and nitric acids, the improvement, is described comprising: after etching of zirconium metallic articles in the bath for a period of time such that the etching rate has diminished from an initial rate to a lesser rate, adding hydrofluoric acid and nitric acid to the exhausted bath to adjust the concentration and ratio of hydrofluoric acid to nitric acid therein to a value substantially that of the initial concentration and ratio and thereby regenerate the etching solution without removal of dissolved zirconium therefrom; and etching further zirconium metallic articles in the regenerated etching bath

Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

Energy Technology Data Exchange (ETDEWEB)

John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

2013-03-01

The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has been evaluated as an acceptable benchmark experiment. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

Internal structure of reactor building for Madras Atomic Power Project

International Nuclear Information System (INIS)

Pandit, D.P.

1975-01-01

The structural configuration and analysis of structural elements of the internal structure of reactor building for the Madras Atomic Power Project has been presented. Two methods of analysis of the internal structure, viz. Equivalent Plane Frame and Finite Element Method, are explained and compared with the use of bending moments obtained. (author)

Zirconium sponge and other forms of virgin metal for nuclear applications - approved standard 1973

International Nuclear Information System (INIS)

Anon.

1975-01-01

This specification covers virgin zirconium metal commonly designated as sponge because of its porous, sponge-like texture, but it may also take other forms such as chunklets. One grade is described which is designated as reactor grade R-1, suitable for use in nuclear applications. The most important characteristic of the reactor grade is its low nuclear cross section as achieved by removal of hafnium and careful quality control in manufacturing procedures to prevent contamination with other high cross section materials

Synthesis of zirconium by zirconium tetrachloride reduction by magnesio-thermia. Experimental study and modelling

International Nuclear Information System (INIS)

Basin, N.

2001-01-01

This work deals with the synthesis of zirconium. The ore is carbo-chlorinated to obtain the tetrachloride which is then purified by selective condensation and extractive distillation. Zirconium tetrachloride is then reduced by magnesium and the pseudo-alloy is obtained according to the global following reaction (Kroll process): ZrCl 4 + 2 Mg = 2 MgCl 2 . By thermodynamics, it has been shown that the volatilization of magnesium chloride and the formation of zirconium sub-chlorides are minimized when the combined effects of temperature and of dilution with argon are limited. With these conditions, the products, essentially zirconium and magnesium chloride, are obtained in equivalence ratio in the magnesio-thermia reaction. The global kinetics of the reduction process has been studied by a thermal gravimetric method. A thermo-balance device has been developed specially for this kinetics study. It runs under a controlled atmosphere and is coupled to a vapor tetrachloride feed unit. The transformation is modelled supposing that the zirconium and magnesium chloride formation result: 1)of the evaporation of magnesium from its liquid phase 2)of the transfer of magnesium and zirconium tetrachloride vapors towards the front of the reaction located in the gaseous phase 3)of the chemical reaction. In the studied conditions, the diffusion is supposed to be the limiting process. The influence of the following parameters: geometry of the reactive zone, temperature, scanning rate of the argon-zirconium tetrachloride mixture, composition of the argon-zirconium tetrachloride mixture has been experimentally studied and confronted with success to the model. (O.M.)

Final report on: Grain size determination in zirconium alloys (IAEA Research Contract No. 6025/Rb.)

International Nuclear Information System (INIS)

Martinez M, E.

1991-12-01

In spite of the amount of research developed the knowledge still is far from complete and in this basis the International Atomic Energy Agency, (IAEA), by means of the Working Group on Water Reactor Fuel Performance and Technology, initiated, in 1990 the Coordinated Research Programme named Grain Size Determination In Zirconium Alloys. Several countries were invited to participate and to contribute to the main objective of the programme, which can be state as: To develop a unified metallographic technique capable to show the microstructure of zircaloy in a reproducible and uniform manner. To fulfill the objective the following goals were established: A. To measure the grain size and perform an statistical treatment, in samples prepared specifically to show different amounts of cold work, recrystallization and grain growth. B. To compare the results obtained by the different laboratories involved in the programme. C. Finally, after the Ugine meeting, also the determination of the recrystallization and grain growth kinetics. (Author)

Shadow corrosion testing in the INCA facility in the Studsvik R2 reactor

International Nuclear Information System (INIS)

Nystrand, A.C.; Lassing, A.

1999-01-01

Shadow corrosion is a phenomenon which occurs when zirconium alloys are in contact with or in proximity to other metallic objects in a boiling water reactor environment (BWR, RBMK, SGHWR etc.). An enhanced corrosion occurs on the zirconium alloy with the appearance of a 'shadow' of the metallic object. The magnitude of the shadow corrosion can be significant, and is potentially limiting for the lifetime of certain zirconium alloy components in BWRs and other reactors with a similar water chemistry. In order to evaluate the suitability of the In-Core Autoclave (INCA) in the Studsvik R2 materials testing reactor as an experimental facility for studying shadow corrosion, a demonstration test has been performed. A number of test specimens consisting of Zircaloy-2 tubing in contact with Inconel were exposed in an oxidising water chemistry. Some of the specimens were placed within the reactor core and some above the core. The conclusion of this experiment after post irradiation examination is that it is possible to use the INCA facility in the Studsvik R2 reactor to develop a significant level of shadow corrosion after only 800 hours of irradiation. (author)

Phase equilibria at the Zirconium metal purification

International Nuclear Information System (INIS)

Dwiretnani-Sudjoko; Busron-Masduki; Sunardjo; Budi-Sulistyo

1996-01-01

It was investigated the research in the purification of zirconium metal, which was results from the reduction process, by adding heat in the vacuum environment. The process was done in batch in the stainless steel reactor, equiped with vacuum pump and electric heater. The investigated variable were process temperature and pressure. From this research it was obtained that equilibrium constant for MgCl 2 and Mg were expressed in the equation K M g C l 2 = 0.9011 P 1 .3779 1.06552 T and K M g = 6.0115P + 1.35256T - 6.93912

Coupling NMR with Synchrotron radiation at high temperature for the study of molten fluorides : applied to zirconium fluorides

International Nuclear Information System (INIS)

Maksoud, Louis

2013-01-01

Molten fluorides are used in Molten Salt Reactors MSR such as the non moderated fast reactor MSFR, where the molten salt LiF-ThF 4 is the fuel and the coolant. The formation of fission products (FP) such as lanthanides, during the reactor operation, possibly modifies the physicochemical properties of the melt. It is therefore important to characterize the melt from the structural and the dynamics point of view in order to determine its properties. Because of problems related to the radioactivity of thorium, as well as requirements related to spectroscopic methods, the system studied in this thesis is the LiF-ZrF 4 -LaF 3 (zirconium and lanthanum are possible FP). The approach followed in this thesis combines measurements by NMR spectroscopy and EXAFS at 850 C with molecular dynamics simulations. In the molten salt, we have shown the existence of zirconium and lanthanum complexes with different coordination numbers, whose proportions depend on the composition. Depending on the content of ZrF 4 , [ZrF 7 ] 3- species are dominant but change slightly and are further connected between each other's via bridging fluorine. The addition of LaF 3 to the mixture stabilizes the 7 coordination number around the zirconium and tends to enrich the environment of lanthanum with fluorides. A medium-range order is established between the various complexes containing zirconium and lanthanum due to bridging fluorine. Species dynamics is slower when the amount of either ZrF 4 or LaF 3 is higher. We noted a significant effect on the structure and dynamics of species starting 10 mol% LaF 3 added to the medium. The data obtained by this novel approach concerning the chemistry of the molten salt in MSR containing FP, are fundamental to improve the separation of these products and optimize the process. (author)

Moltex Energy's stable salt reactors

International Nuclear Information System (INIS)

O'Sullivan, R.; Laurie, J.

2016-01-01

A stable salt reactor is a molten salt reactor in which the molten fuel salt is contained in fuel rods. This concept was invented in 1951 and re-discovered and improved recently by Moltex Energy Company. The main advantage of using molten salt fuel is that the 2 problematic fission products cesium and iodine do not exist in gaseous form but rather in a form of a salt that present no danger in case of accident. Another advantage is the strongly negative temperature coefficient for reactivity which means the reactor self-regulates. The feasibility studies have been performed on a molten salt fuel composed of sodium chloride and plutonium/uranium/lanthanide/actinide trichloride. The coolant fluid is a mix of sodium and zirconium fluoride salts that will need low flow rates. The addition of 1 mol% of metal zirconium to the coolant fluid reduces the risk of corrosion with standard steels and the addition of 2% of hafnium reduces the neutron dose. The temperature of the coolant is expected to reach 650 Celsius degrees at the exit of the core. This reactor is designed to be modular and it will be able to burn actinides. (A.C.)

Science Hall of Atomic Energy in Research Reactor Institute, Kyoto University

International Nuclear Information System (INIS)

Hayashi, Takeo

1979-01-01

The Science Hall of Atomic Energy was built as a subsidiary facility of the Research Reactor Institute, Kyoto University. The purpose of this facility is to accept outside demands concerning the application of the research reactor. The building is a two story building, and has the floor area of 901.47 m 2 . There are an exhibition room, a library, and a big lecture room. In the exhibition room, models of the Kyoto University Research Reactor and the Kyoto University Critical Assembly are placed. Various pictures concerning the application of the reactor are on the wall. In the library, people from outside of the Institute can use various books on science. Books for boys and girls are also stocked and used for public use. At the lecture room, various kinds of meeting can be held. (Kato, T.)

Chemistry, spectroscopy and isotope separation of zirconium and its compounds as revealed by laser diagnostics of laser produced metal beams

International Nuclear Information System (INIS)

Hackett, P.A.; Humphries, M.; Rayner, D.M.; Bourne, O.L.; Mitchell, A.

1986-01-01

Recent work from the author's laboratory on zirconium beams is reviewed. Zirconium metal beams have been produced by laser vaporization of solid zirconium targets coupled with supersonic expansion of helium gas. The resultant supersonic metal beam is shown to present an ideal environment for various spectroscopic techniques. The state distribution of zirconium atoms in the beam is obtained from low resolution laser induced fluorescence (LIF) studies. High resolution LIF studies give information on the hyperfine splitting in the ground state of the zirconium-91 isotope. Information on the hyperfine splitting in the excited state is obtained from quantum beat spectroscopy. Low resolution 2 color multiphoton ionization spectroscopy using a XeCl laser allows isotope separation of all isotopes of zirconium. These metal beams are highly reactive and can be used to produce novel chemical species. The results of two studies in which a reactant is added to the expansion gas are reported here. Zirconium oxide (ZrO), a molecule observed in the emission spectra of cool stars and in laboratory studies at high temperatures, is produced in a low temperature, collision free environment by adding small quantities of oxygen to the expansion gas. Zirconium fluoride (ZrF), a molecule previously unobserved, is produced by the addition of small quantities of CF/sub 4/

Performance of U-Pu-Zr fuel cast into zirconium molds

International Nuclear Information System (INIS)

Crawford, D.C.; Lahm, C.E.; Tsai, H.

1992-01-01

Current fabrication techniques for the integral fast reactor (IFR) fuel utilize injection casting into quartz molds after reprocessing in the IFR fuel cycle facility. The quartz molds are destroyed during the fuel demolding process, and the quartz residue must therefore be treated as contaminated waste. Alternatively, if the fuel can be cast into molds that remain as part of the fuel slugs (i.e., if the fuel can be left inside the molds for irradiation), then the quartz mold contribution to the waste stream can be eliminated. This possibility is being addresssed in an ongoing effort to evaluate the irradiation performance of fuel cast into zirconium sheaths rather than quartz molds. Zirconium was chosen as the sheath material because it is the component of the U-Pu-Zr fuel alloy that raises the alloy solidus temperatures and provides resistance to fuel-cladding chemical interaction (FCCI)

Neutron activation of chlorine in zirconium and zirconium alloys use of the matrix as comparator

International Nuclear Information System (INIS)

Cohen, I.M.; Gomez, C.D.; Mila, M.I.

1981-01-01

A procedure is described for neutron activation analysis of chlorine in zirconium and zirconium alloys. Calculation of chlorine concentration is performed relative to zirconium concentration in the matrix in order to minimize effects of differences in irradiation and counting geometry. Principles of the method and the results obtained are discussed. (author)

The selectivity of zirconium phosphate for caesium in electrochemical ion exchange

International Nuclear Information System (INIS)

Lain, M.J.

1988-11-01

The properties of amorphous zirconium phosphate are investigated as an inorganic ion exchanger for use in liquid waste treatment by electrochemical ion exchange. Experiments to determine and increase the selectivity for caesium exchange over sodium are discussed, including various pulsed waveforms and studies with rotating membranes. Automation of a sampling system with pH and atomic absorption measurements is described. (author)

Protection of zirconium and its alloys by metallic coatings

International Nuclear Information System (INIS)

Loriers, H.; Lafon, A.; Darras, R.; Baque, P.

1968-01-01

At 600 deg. C in an atmosphere of carbon dioxide, zirconium and its alloys undergo corrosion which presents two aspects simultaneously: - formation of a surface layer of zirconia, - dissolution of oxygen in the alloy sub-layer leading to brittleness. The two phenomena greatly restrict the possibilities of using zirconium alloys as a canning material for fuel elements in CO 2 cooled nuclear reactors. An attempt has thus been made to limit, and perhaps to suppress, the corrosion effects in zirconium under these conditions by protecting it with metallic coatings. A first attempt to obtain a protection using copper-based coatings did not produce the result hoped for. Aluminium coatings produced by vacuum evaporation, followed by a consolidating thermal treatment make it possible to prevent the formation of the zirconia layer, but they do not eliminate the hardening effect produced by oxygen diffusion. On the other hand, electrolytically produced chromium deposits whose adherence is improved by a thermal vacuum treatment, counteract both these phenomena simultaneously. A similar result has been obtained with coatings of molybdenum produced by the technique of high-frequency inductive plasma sputtering. The particular effectiveness of the last two types of coatings is due to their structures characterized by the existence of an adherent film of chromium or molybdenum in the free state. (authors) [fr

Passivation of mechanically polished, chemically etched and anodized zirconium in various aqueous solutions: Impedance measurements

International Nuclear Information System (INIS)

Abo-Elenien, G.M.; Abdel-Salam, O.E.

1987-01-01

Zirconium and its alloys are finding increasing applications especially in water-cooled nuclear reactors. Because of the fact that zirconium is electronegative (E 0 = -1.529V) its corrosion resistance in aqueous solutions is largely determined by the existence of a thin oxide film on its surface. The structure and properties of this film depend in the first place on the method of surface pre-treatment. This paper presents an experimental study of the nature of the oxide film on mechanically polished, chemically etched and anodized zirconium. Ac impedance measurements carried out in various acidic, neutral and alkaline solutions show that the film thickness depends on the method of surface pre-treatment and the type of electrolyte solution. The variation of the potential and impedance during anodization of zirconium at low current density indicates that the initial stages of polarization consist of oxide build-up at a rate dependent on the nature of the electrode surface and the electrolyte. Oxygen evolution commences at a stage where oxide thickening starts to decline. The effect of frequency on the measured impedance indicates that the surface reactivity, and hence the corrosion rate, decreases in the following order: mechanically polished > chemically etched > anodized

Extraction of zirconium from raffinate stream of Zirconium Oxide Plant raffinate

International Nuclear Information System (INIS)

Pandey, Garima; Chinchale, R.; Renjith, A.U.; Mukhopadhyay, S.; Shenoy, K.T.; Ghosh, S.K.

2013-01-01

Recovery of metals from dilute streams is a major task in nuclear industry in the view of environmental remediation and value recovery. Presently solvent extraction process is employed on the commercial scale to recover nuclear pure zirconium using TBP as extractant. The waste stream of TBP extraction process contains about 1.2 gpl of Zirconium in nitrate form. At present there is no process to recover Zirconium from this raffinate stream. Hence, under the present study recovery of zirconium from the raffinate stream of Zirconium Oxide Plant Raffinate has been investigated. TBP, which is the most commonly used solvent in the nuclear industry is not suitable for the extraction of zirconium from lean solution at low acidity as its distribution coefficient is less than one. In search of a suitable extractant Mixed Alkyl Phosphine Oxide (MAPO) was investigated as potential carrier. Parametric batch studies for various equilibrium data like extractant concentration, strippant concentration, solvent reusability, equilibration time, acidity etc. were done to optimize the process condition. For the distribution studies, equal volumes of the raffinate and organic phase were shaken at room temperature in digital wrist action shaker for 10 minutes to ensure complete equilibrium. It was found that 0.1 M MAPO in 80:20 dodecane: isodecanol is suitable for extraction of Zr at 2 N acidity. 0.1 M MAPO gives distribution coefficient in the range of 12-15 for Zr. The slope of log-log plot between MAPO concentration and K, suggests involvement of 3 molecules of MAPO in the formation of extracting species. 0.2 M Oxalic acid was able to completely back extract Zr from the organic phase into aqueous phase. Also good regeneration capacity of MAPO projects its potential to be used as extractant for the process. Based on the equilibrium studies, Dispersion Liquid Membrane configuration in hollow fiber contactor was explored for the extraction of Zirconium from Zirconium Nitrate Pure

Molecular dynamics simulations of irradiation cascades in alpha-zirconium under macroscopic strain

Energy Technology Data Exchange (ETDEWEB)

Di, Sali [Department of Mechanical and Materials Engineering, Queen’s University, Kingston, ON, Canada K7L 3N6 (Canada); Yao, Zhongwen, E-mail: yaoz@me.queensu.ca [Department of Mechanical and Materials Engineering, Queen’s University, Kingston, ON, Canada K7L 3N6 (Canada); Daymond, Mark R. [Department of Mechanical and Materials Engineering, Queen’s University, Kingston, ON, Canada K7L 3N6 (Canada); Gao, Fei [Pacific Northwest National Laboratory, Richland, WA 99352 (United States)

2013-05-15

Numerous computer simulation studies have been performed on the radiation damage of zirconium. In contrast to most of the work in the literature which has focused on the effects of temperature and recoil energy on defect production and defect clustering, we have developed a computational model to consider the influence of elastic strain field on the formation of defects and their clusters, as strain is commonly present in a real reactor environment. In this work, irradiation induced displacement cascades in alpha-zirconium experiencing a macroscopic strain have been studied by molecular dynamics (MD) simulations using a many-body interatomic potential. The external strain mainly affects the size of defect clusters rather than the total number of defects. The sizes of interstitial and vacancy clusters respond differently to the external strain conditions.

Report of researches by common utilization of facilities in Kyoto University Research Reactor Institute, first half of fiscal year 1981

International Nuclear Information System (INIS)

1983-01-01

The technical report of the Kyoto University Research Reactor Institute is published any time to immediately report on the results of the functional tests of various experimental facilities, the test results for the products made for trial, radiation control, the situation of waste treatment, the data required for research and experiment such as the reports of study meetings, the conspicuous results obtained amid researches, new processes, and the discussion on other papers and reports. In this report, the title, the names of reporters and the summary of 57 researches carried out by the common utilization of the facilities in the Kyoto University Research Reactor Institute are collected. The themes of the researches are such as neutron radiography using a research reactor, measurement of Zr/Hf ratio in zirconium, interstitial germanium atoms in thermal neutron irradiation study, measurement of induced radioactivity due to neutrons in Nagasaki and Hiroshima atomic bombings, properties of semiconductor electrons in radiation study, induction of mutation in crops by neutron irradiation and utilization for breeding, thermal fluorescence mechanism of alkali halide and MgO single crystals, atomic configuration in PZT rhombohedron phase, modulated structure of Cu-Co alloys, excitation of nuclei by positron annihilation and others. (Kako, I.)

A contribution to the study of the mixed uranium-plutonium mono-carbides containing small quantities of zirconium

International Nuclear Information System (INIS)

Bocker, S.

1970-03-01

We have studied a mixed monocarbide, type (U,Pu)C, containing small additions of zirconium for the application as a fast neutron reactor fuel. A preliminary study was conducted on the (U,Zr)C monocarbide (Report CEA-R-3765(1). It was found that small additions of zirconium to the uranium-plutonium monocarbide improve a number of properties such as atmospheric corrosion, the hardness, and particularly the compatibility with 316 stainless steel. However, properties such as the coefficient of expansion and the melting point are only slightly changed. The relative percentage of Pu/U+Pu in the monocarbide was fixed at 20 per cent. Two processes of fabrication were employed: casting in an arc furnace, sintering, carried out after having the hydrides of the metals carburized. The metallurgical results indicate, that the above mentioned fuel might be of interest for fast neutron reactor application. (author) [fr

Modification in band gap of zirconium complexes

Energy Technology Data Exchange (ETDEWEB)

Sharma, Mayank, E-mail: mayank30134@gmail.com; Singh, J.; Chouhan, S. [Department of Physics, ISLE, IPS Academy, Indore (M.P.) (India); Mishra, A. [School of Physics, Devi Ahilya Vishwavidyalaya, Indore (M.P.) (India); Shrivastava, B. D. [Govt. P. G. College, Biora (M.P.) (India)

2016-05-06

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.

Chemical aspects of hydrogen ingress in zirconium and zircaloy pressure tubes: ageing management of Indian PHWR coolant channels - determination of hydrogen and deuterium

International Nuclear Information System (INIS)

Sayi, Y.S.; Shankaran, P.S.; Yadav, C.S.; Ramanjaneyulu, P.S.; Venugopal, V.; Ramakumar, K.L.; Chhapru, G.C.; Prasad, R.; Jain, H.C.; Sood, D.D.

2009-02-01

Pressurized heavy water reactors (PHWRs) use zirconium and zirconium based alloys as clad and coolant tubes since its beginning. The first ever zircaloy-2 pressure tube failure occurred in 1983 at Ontario Hydro's Pickering Unit 2 in Canada which necessitated a thorough examination of causes of such failure. The failure was attributed to massive hydriding at the failed spot of pressure tube. Continuous usage of zirconium alloys could result in their hydrogen and deuterium pick-up leading to hydrogen/ deuterium embrittlement. The life of the zircaloy coolant channels is dictated by hydrogen/deuterium content and hence ageing management of the pressure tubes is essential for ensuring their trouble-free usage. It is desirable to have a sound knowledge on the chemical aspects of zirconium and zirconium based alloys metallurgy, the mechanistic principles of hydrogen ingress into the pressure tubes during in reactor service, and identifying suitable analytical methodologies for precise and accurate determination of hydrogen in wafer thin sliver samples carved out from insides of pressure tubes without causing any structural damage so that it can continue to remain in service. This is desirable so that the ageing management does not result in cost-escalation. This report is divided in to three main parts. The first part deals with the chemical aspects of zirconium and zirconium based alloy metallurgy, the mechanism of hydrogen pick-up and hydride formation in zirconium matrix. The second part describes various methodologies and their limitations, available for hydrogen/deuterium determination. The third part deals in detail, about the extensive investigations carried out at Radioanalytical Chemistry Division (RACD) in Radiochemistry and Isotope Group for establishing an indigenously developed hot vacuum extraction system in combination with quadrupole mass spectrometry for precise determination of hydrogen and deuterium in wafer thin sliver sample of zircaloy. The

The Effect of Hydrogen and Hydrides on the Integrity of Zirconium Alloy Components Delayed Hydride Cracking

CERN Document Server

Puls, Manfred P

2012-01-01

By drawing together the current theoretical and experimental understanding of the phenomena of delayed hydride cracking (DHC) in zirconium alloys, The Effect of Hydrogen and Hydrides on the Integrity of Zirconium Alloy Components: Delayed Hydride Cracking provides a detailed explanation focusing on the properties of hydrogen and hydrides in these alloys. Whilst the focus lies on zirconium alloys, the combination of both the empirical and mechanistic approaches creates a solid understanding that can also be applied to other hydride forming metals.   This up-to-date reference focuses on documented research surrounding DHC, including current methodologies for design and assessment of the results of periodic in-service inspections of pressure tubes in nuclear reactors. Emphasis is placed on showing that our understanding of DHC is supported by progress across a broad range of fields. These include hysteresis associated with first-order phase transformations; phase relationships in coherent crystalline metallic...

Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

Energy Technology Data Exchange (ETDEWEB)

John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

2011-03-01

The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

Energy Technology Data Exchange (ETDEWEB)

Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Maddock, Thomas L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Marshall, Margaret A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Ning [Idaho National Lab. (INL), Idaho Falls, ID (United States); Phillips, Ann Marie [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schreck, Kenneth A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Briggs, J. Blair [Idaho National Lab. (INL), Idaho Falls, ID (United States); Woolstenhulme, Eric W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bolin, John M. [General Atomics, San Diego, CA (United States); Veca, Anthony [General Atomics, San Diego, CA (United States); McKnight, Richard D. [Argonne National Lab. (ANL), Argonne, IL (United States); Lell, Richard M. [Argonne National Lab. (ANL), Argonne, IL (United States)

2014-03-01

The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

Plasma arc melting of zirconium

International Nuclear Information System (INIS)

Tubesing, P.K.; Korzekwa, D.R.; Dunn, P.S.

1997-01-01

Zirconium, like some other refractory metals, has an undesirable sensitivity to interstitials such as oxygen. Traditionally, zirconium is processed by electron beam melting to maintain minimum interstitial contamination. Electron beam melted zirconium, however, does not respond positively to mechanical processing due to its large grain size. The authors undertook a study to determine if plasma arc melting (PAM) technology could be utilized to maintain low interstitial concentrations and improve the response of zirconium to subsequent mechanical processing. The PAM process enabled them to control and maintain low interstitial levels of oxygen and carbon, produce a more favorable grain structure, and with supplementary off-gassing, improve the response to mechanical forming

Solvent extraction of titanium(IV), zirconium(IV) and hafnium(IV) salicylates using liquid ion exchangers

Energy Technology Data Exchange (ETDEWEB)

Sundaramurthi, N M; Shinde, V M

1989-02-01

A solvent extraction method is proposed for the extraction of quadrivalent titanium, zirconium an hafnium from salicylate media using liquid ion exchangers such as Aliquat 336 and trioctylamine dissolved in xylene. The optimum conditions were evaluated from a critical study of the following: pH, salicylate concentration, amine concentration, diluent and period of equilibration. The method allows the separation of titanium, zirconium and hafnium from binary mixtures containing commonly associated metal ions and is applicable to the analysis of real samples such as BCS-CRM 387 nimonic 901, BCS-CRM 243/4 ferro-titanium, BCS-CRM 307 magnesium alloy and BCS-CRM 388 zircon. Titanium is determined either with hydrogen peroxide or by atomic absorption spectrometry whereas zirconium and hafnium are determined spectrophotometrically with Alizarin Red S and Zylenol Orange, respectively. The results of both separation and analysis are reported. The method is precise, accurate and fast.

Zirconium behaviour in purex process solutions

International Nuclear Information System (INIS)

Shu, J.

1982-01-01

The extraction behaviour of zirconium, as fission product, in TBP/diluent- HNO 3 -H 2 O systems, simulating Purex solutions, is studied. The main purpose is to attain an increasing in the zirconium decontamination factor by adjusting the extraction parameters. Equilibrium diagram, TBP concentration, aqueous:organic ratio, salting-out effects and, uranium loading in the organic phase were the main factors studied. All these experiments had been made with zirconium in the 10 - 2 - 10 - 3 concentration range. The extractant degradation products influence uppon the zirconium behaviour was also verified. With the obtained data it was possible to introduce some modification in the standard Purex flow-sheet in order to obtain the uranium product with higher zirconium decontamination. (Author) [pt

Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station (LWBR Development Program)

International Nuclear Information System (INIS)

Massimino, R.J.; Williams, D.A.

1983-05-01

This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core

Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station (LWBR Development Program)

Energy Technology Data Exchange (ETDEWEB)

Massimino, R.J.; Williams, D.A.

1983-05-01

This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core.

Corrosion performance of new Zircaloy-2-based alloys

International Nuclear Information System (INIS)

Rudling, P.; Mikes-Lindbaeck, M.; Lethinen, B.; Andren, H.O.; Stiller, K.

1994-01-01

A material development project was initiated to develop a new zirconium alloy, outside the ASTM specifications for Zircaloy-2 and Zircaloy-4, with optimized hydriding and corrosion properties for both boiling water reactors and pressurized water reactors. A number of different alloys were manufactured. These alloys were long-term corrosion tested in autoclaves at 400 C in steam. Also, a 520 C/24 h steam test was carried out. The zirconium metal microstructure and the chemistry of precipitates were characterized by analytical electron microscopy. The metal matrix chemistry was determined by atom probe analysis. The paper describes the correlations between corrosion material performance and zirconium alloy microstructure

Change in plan for installation of nuclear reactor in No.1 atomic powered vessel of Japan Atomic Energy Research Institute (change in purpose of use and in method for nuclear reactor installation and spent fuel disposal) (report)

International Nuclear Information System (INIS)

1987-01-01

This report, compiled by the Nuclear Safety Commission to be submitted to the Prime Minister, deals with studies concerning some changes in the plan for the installation of a nuclear reactor in the No.1 atomic powered vessel to be constructed under the Japan Atomic Energy Research Institute (changes in the purpose of its use and in the methods for the nuclear reactor installation and spent fuel disposal). The conclusions of and procedures for the examination and evaluation are presented and then detailes of the studies are described. The study on the location requirements for the incidental land facilities at Sekinehama covers various conditions concerning the location, geology, earthquakes, meteorology, hydrology and social environment. The study on the safety design of the nuclear reactor facilities deals with the reactor, fuel handling facilities and other auxiliary facilities, as well as various land facilities to be constructed at Sekinehama including the reactor facilities and other facilities for fuel handling, waste disposal and protection and management of radioactive rays. Evaluation of possible radiation emission is shown and the accident analysis is also addressed. (Nogami, K.)

Fine-grained zirconium-base material

Science.gov (United States)

Van Houten, G.R.

1974-01-01

A method is described for making zirconium with inhibited grain growth characteristics, by the process of vacuum melting the zirconium, adding 0.3 to 0.5% carbon, stirring, homogenizing, and cooling. (Official Gazette)

Components made of corrosion-resistent zirconium alloy and method for its production

International Nuclear Information System (INIS)

Hanneman, R.E.; Urquhart, A.W.; Vermilyea, D.A.

1977-01-01

The invention deals with a method to increase the resistance of zirconium alloys to blister corrosion which mainly occurs in boiling-water nuclear reactors. According to the method described, the surface of the alloy body is coated with a thin film of a suitable electronically conducting material. Gold, silver, platinum, nickel, chromium, iron and niobium are suitable as coating materials. The invention is more closely explained by means of examples. (GSC) [de

Microhardness and microplasticity of zirconium nitride

International Nuclear Information System (INIS)

Neshpor, V.S.; Eron'yan, M.A.; Petrov, A.N.; Kravchik, A.E.

1978-01-01

To experimentally check the concentration dependence of microhardness of 4 group nitrides, microhardness of zirconium nitride compact samples was measured. The samples were obtained either by bulk saturation of zirconium iodide plates or by chemical precipitation from gas. As nitrogen content decreased within the limits of homogeneity of zirconium nitride samples where the concentration of admixed oxygen was low, the microhardness grew from 1500+-100 kg/mm 2 for ZrNsub(1.0) to 27000+-100 kg/mm 2 for ZrNsub(0.78). Microplasticity of zirconium nitride (resistance to fracture) decreased, as the concentration of nitrogen vacancies was growing

Analysis of zirconium and nickel based alloys and zirconium oxides by relative and internal monostandard neutron activation analysis methods

International Nuclear Information System (INIS)

Shinde, Amol D.; Acharya, Raghunath; Reddy, Annareddy V. R.

2017-01-01

The chemical characterization of metallic alloys and oxides is conventionally carried out by wet chemical analytical methods and/or instrumental methods. Instrumental neutron activation analysis (INAA) is capable of analyzing samples nondestructively. As a part of a chemical quality control exercise, Zircaloys 2 and 4, nimonic alloy, and zirconium oxide samples were analyzed by two INAA methods. The samples of alloys and oxides were also analyzed by inductively coupled plasma optical emission spectroscopy (ICP-OES) and direct current Arc OES methods, respectively, for quality assurance purposes. The samples are important in various fields including nuclear technology. Samples were neutron irradiated using nuclear reactors, and the radioactive assay was carried out using high-resolution gamma-ray spectrometry. Major to trace mass fractions were determined using both relative and internal monostandard (IM) NAA methods as well as OES methods. In the case of alloys, compositional analyses as well as concentrations of some trace elements were determined, whereas in the case of zirconium oxides, six trace elements were determined. For method validation, British Chemical Standard (BCS)-certified reference material 310/1 (a nimonic alloy) was analyzed using both relative INAA and IM-NAA methods. The results showed that IM-NAA and relative INAA methods can be used for nondestructive chemical quality control of alloys and oxide samples

Analysis of zirconium and nickel based alloys and zirconium oxides by relative and internal monostandard neutron activation analysis methods

Energy Technology Data Exchange (ETDEWEB)

Shinde, Amol D.; Acharya, Raghunath; Reddy, Annareddy V. R. [Bhabha Atomic Research Centre, Mumbai (India)

2017-04-15

The chemical characterization of metallic alloys and oxides is conventionally carried out by wet chemical analytical methods and/or instrumental methods. Instrumental neutron activation analysis (INAA) is capable of analyzing samples nondestructively. As a part of a chemical quality control exercise, Zircaloys 2 and 4, nimonic alloy, and zirconium oxide samples were analyzed by two INAA methods. The samples of alloys and oxides were also analyzed by inductively coupled plasma optical emission spectroscopy (ICP-OES) and direct current Arc OES methods, respectively, for quality assurance purposes. The samples are important in various fields including nuclear technology. Samples were neutron irradiated using nuclear reactors, and the radioactive assay was carried out using high-resolution gamma-ray spectrometry. Major to trace mass fractions were determined using both relative and internal monostandard (IM) NAA methods as well as OES methods. In the case of alloys, compositional analyses as well as concentrations of some trace elements were determined, whereas in the case of zirconium oxides, six trace elements were determined. For method validation, British Chemical Standard (BCS)-certified reference material 310/1 (a nimonic alloy) was analyzed using both relative INAA and IM-NAA methods. The results showed that IM-NAA and relative INAA methods can be used for nondestructive chemical quality control of alloys and oxide samples.

SEPARATION OF HAFNIUM FROM ZIRCONIUM

Science.gov (United States)

Overholser, L.B.; Barton, C.J. Sr.; Ramsey, J.W.

1960-05-31

The separation of hafnium impurities from zirconium can be accomplished by means of organic solvent extraction. The hafnium-containing zirconium feed material is dissolved in an aqueous chloride solution and the resulting solution is contacted with an organic hexone phase, with at least one of the phases containing thiocyanate. The hafnium is extracted into the organic phase while zirconium remains in the aqueous phase. Further recovery of zirconium is effected by stripping the onganic phase with a hydrochloric acid solution and commingling the resulting strip solution with the aqueous feed solution. Hexone is recovered and recycled by means of scrubbing the onganic phase with a sulfuric acid solution to remove the hafnium, and thiocyanate is recovered and recycled by means of neutralizing the effluent streams to obtain ammonium thiocyanate.

Zirconium

Science.gov (United States)

Bedinger, G.M.

2013-01-01

Zirconium is the 20th most abundant element in the Earth’s crust. It occurs in a variety of rock types and geologic environments but most often in igneous rocks in the form of zircon (ZrSiO4). Zircon is recovered as a coproduct of the mining and processing of heavy mineral sands for the titanium minerals ilmenite and rutile. The sands are formed by the weathering and erosion of rock containing zircon and titanium heavy minerals and their subsequent concentration in sedimentary systems, particularly in coastal environments. A small quantity of zirconium, less than 10 kt/a (11,000 stpy), compared with total world production of 1.4 Mt (1.5 million st) in 2012, was derived from the mineral baddeleyite (ZrO2), produced from a single source in Kovdor, Russia.

MULTILAYER COMPOSITE PLASMA COATINGS ON SCREEN PROTECTION ELEMENTS BASED ON ZIRCONIUM DIOXIDE

Directory of Open Access Journals (Sweden)

V. A. Okovity

2017-01-01

Full Text Available The paper contains results of investigations pertaining to an influence of plasma jet parameters (current, spraying distance, consumption of plasma formation gas (nitrogen, fractional composition of initial powder and degree of cooling with compressed air on anti-meteoric coating characteristics. Optimum modes (arc current 600 A; spray distance of 110 mm; consumption of plasma formation gas (nitrogen – 50 l/min; fractional composition of zirconium dioxide powder <50 μm; compressed air consumption for cooling – 1 m3/min; p = 4 bar make it possible to obtain anti-meteoric coatings based on zirconium dioxide with material utilization rate of 62 %, total ceramic layer porosity of 6 %. After exposure of compression plasma flows on a coating in the nitrogen atmosphere a cubic modification of zirconium oxide is considered as the main phase being present in the coating. The lattice parameter of cubic zirconium oxide modification is equal to 0.5174 nm. Taking into consideration usage of nitrogen as plasma formation substance its interaction with zirconium coating atoms occurs and zirconium nitride (ZrN is formed with a cubic crystal lattice (lattice parameter 0.4580 nm. Melting of pre-surface layer takes place and a depth of the melted layer is about 8 μm according to the results of a scanning electron microscopy. Pre-surface layer being crystallized after exposure to compression plasma flows is characterized by a homogeneous distribution of ele-ments and absence of pores formed in the process of coating formation. The coating structure is represented by a set of lar- ge (5–7 μm and small (1–2 μm zirconium oxide particles sintered against each other. Melting of coating surface layer and speed crystallization occur after the impact of compression plasma flows on the formed coating. Cracking of the surface layer arises due to origination of internal mechanical stresses in the crystallized part. While using a scanning electron microscopy a

Solvent extraction of zirconium

International Nuclear Information System (INIS)

Kim, S.S.; Yoon, J.H.

1981-01-01

The extraction of zirconium(VI) from an aqueous solution of constant ionic strength with versatic acid-10 dissolved in benzen was studied as a function of pH and the concentration of zirconium(VI) and organic acid. The effects of sulphate and chlorine ions on the extraction of the zirconium(VI) were briefly examined. It was revealed that (ZrOR 2 .2RH) is the predominant species of extracted zirconium(VI) in the versatic acid-10. The chemical equation and the apparent equilibrium constants thereof have been determined as follows. (ZrOsup(2+))aq+ 2(R 2 H 2 )sub(org) = (ZrOR 2 .2RH)sub(org)+2(H + )aq Ksub(Zr) = (ZrOR 2 .2RH)sub(org)(H + ) 2 /(ZrOsup(2+))sub(aq)(R 2 H 2 )sup(2)sub(org) = 3.3 x 10 -7 . The synergistic effects of TBP and D2EHPA were also studied. In the mixed solvent with 0.1M TBP, the synergistic effect was observed, while the mixed solvent with D2EHPA showed the antisynergistic effect. (Author)

Compositional characterization of hafnium alloy used as control rod material in nuclear reactor

International Nuclear Information System (INIS)

Sharma, P.K.; Bassan, M.K.T.; Avhad, D.K.; Singhal, R.K.

2014-01-01

Hafnium (Hf) is a heavy, steel-gray metal in the reactive metals group that is very closely related to zirconium (Zr) and forms a continuous solid-solution at all concentrations of zirconium and hafnium. Hafnium occurs naturally with zirconium at a ratio of approximately 1:50 and is produced exclusively as a co-product of nuclear-grade zirconium. It is used in a variety of applications where few substitutes are available. Thus with its relatively high thermal neutron absorption cross-section, hafnium's biggest application is as control rod material in nuclear reactors. During this work, major (Zirconium (Zr), Cobalt (Co) and Molybdenum (Mo)) and trace ((Iron (Fe), Nickel (Ni) and Titanium (Ti)) elements were measured in the bulk matrices of Hf. These materials are also associated with other impurities such as O, N, H etc.

Purification of zirconium concentrates

International Nuclear Information System (INIS)

Brown, A.E.P.

1976-01-01

A commercial grade ZrO 2 and an ammonium uranate (yellow cake) are obtained from the caldasito ore processing. This ore is found in the Pocos de Caldas Plateau, State of Minas Gerais, Brazil. Caldasito is an uranigerous zirconium ore, a mixture of zircon and baddeleyite and contains 60% ZrO 2 and 0,3% U 3 O 8 . The chemical opening of the ore was made by alkaline fusion with NaOH at controlled temperature. The zirconium-uranium separation took place by a continuous liquid-liquid extraction in TBP-varsol-HNO 3 -H 2 O system. The raffinate containing zirconium + impurities (aluminium, iron and titanium) was purified by an ion exchange operation using a strong cationic resin [pt

Plasma sprayed and electrospark deposited zirconium metal diffusion barrier coatings

International Nuclear Information System (INIS)

Hollis, Kendall J.; Pena, Maria I.

2010-01-01

Zirconium metal coatings applied by plasma spraying and electrospark deposition (ESD) have been investigated for use as diffusion barrier coatings on low enrichment uranium fuel for research nuclear reactors. The coatings have been applied to both stainless steel as a surrogate and to simulated nuclear fuel uranium-molybdenum alloy substrates. Deposition parameter development accompanied by coating characterization has been performed. The structure of the plasma sprayed coating was shown to vary with transferred arc current during deposition. The structure of ESD coatings was shown to vary with the capacitance of the deposition equipment.

Partial thorium loading in the initial core of Kakrapar atomic power reactor

International Nuclear Information System (INIS)

Balakrishnan, M.R.

1993-01-01

The first unit of Kakrapar nuclear power station has gone critical with some thorium oxide fuel bundles loaded in its core. The thorium helps to flatten the power by reducing neutron flux in the centre of the reactor. However, the placing of the thorium had to be planned with care, because if the neutron flux at a point where a safety rod is located is depressed, the reactivity worth of the safety rod gets reduced. Using a dynamic programing approach, the Reactor Engineering Division of Bhabha Atomic Research Centre worked out a satisfactory configuration for loading the thorium bundles

Quantitative analysis of nickel in zirconium and zircaloy; Dosage du nickel dans le zirconium et dans le zircaloy

Energy Technology Data Exchange (ETDEWEB)

Rastoix, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1957-07-01

A rapid spectrophotometric has been developed for determination of 10 to 1000 ppm of Ni in zirconium and zircaloy using dimethylglyoxime. Iron, copper, tin and chromium, do not interfere at the concentration usually present in zirconium and its alloys. (author) [French] On determine colorimetriquenent 10 a 1000 ppm de Ni dans le zirconium et le zircaloy par photo colorimetrie a 440 m{mu} de la dimethylglyoxime nickelique. Le dosage est rapide. Le fer, le cuivre, l'etain, le chrome ne genent pas aux concentrations habituellement rencontrees dans le zirconium et ses alliages. (auteur)

40 CFR 721.9973 - Zirconium dichlorides (generic).

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 30 2010-07-01 2010-07-01 false Zirconium dichlorides (generic). 721... Substances § 721.9973 Zirconium dichlorides (generic). (a) Chemical substance and significant new uses subject to reporting. (1) The chemical substances identified generically as zirconium dichlorides (PMNs P...

Molten salt scrubbing of zirconium or hafnium tetrachloride

International Nuclear Information System (INIS)

Lee, E.D.; McLaughlin, D.F.

1990-01-01

This patent describes a continuous process for removing impurities of iron or aluminum chloride or both from vaporous zirconium or hafnium chloride or both. It comprises: introducing impure zirconium or hafnium chloride vapor or both into a middle portion of an absorbing column containing a molten salt phase, the molten salt phase absorbing the impurities of iron or aluminum chloride or both to produce chloride vapor stripped of zirconium or hafnium chloride; introducing sodium or potassium chloride or both into a top portion of the column; controlling the top portion of the column to between 300--375 degrees C.; heating a bottom portion of the column to 450--550 degrees C. To vaporize zirconium chloride or hafnium chloride or hafnium and zirconium chloride from the molten salt; withdrawing molten salt substantially free of zirconium and hafnium chloride from the bottom portion of the column; and withdrawing zirconium chloride or hafnium chloride or hafnium and zirconium chloride vapor substantially free of impurities of iron and aluminum chloride from the top of the column

The low-temperature water-water reactor for district heating atomic power plant (DHPP)

International Nuclear Information System (INIS)

Skvortsov, S.A.; Sokolov, I.N.; Krauze, L.V.; Nikiporetz, Yu.G.; Philimonov, Yu.V.

1977-01-01

The district heating atomic power plant in the article is distinguished by the increased reliability and safety of operation that was provided by the use of following main principles: relatively low parameters of the coolant; the intergral arrangement of equipment and accordingly the minimum branching of the reactor circuit; the natural circulation of coolant of the primary circuit in the steady-state, transient and emergency regimes of reactor operation; the considerable reserves of cold water of the primary circuit in the reactor vessel, providing the emergency cooling; the application of two shells each of which is designed for the total working pressure, the second shell is made of prestressed reinforced concrete that eliminates its brittle failure. (M.S.)

Experimental approach and micro-mechanical modeling of the mechanical behavior of irradiated zirconium alloys; Approche experimentale et modelisation micromecanique du comportement des alliages de zirconium irradies

Energy Technology Data Exchange (ETDEWEB)

Onimus, F

2003-12-01

Zirconium alloys cladding tubes containing nuclear fuel of the Pressurized Water Reactors constitute the first safety barrier against the dissemination of radioactive elements. Thus, it is essential to predict the mechanical behavior of the material in-reactor conditions. This study aims, on the one hand, to identify and characterize the mechanisms of the plastic deformation of irradiated zirconium alloys and, on the other hand, to propose a micro-mechanical modeling based on these mechanisms. The experimental analysis shows that, for the irradiated material, the plastic deformation occurs by dislocation channeling. For transverse tensile test and internal pressure test this channeling occurs in the basal planes. However, for axial tensile test, the study revealed that the plastic deformation also occurs by channeling but in the prismatic and pyramidal planes. In addition, the study of the macroscopic mechanical behavior, compared to the deformation mechanisms observed by TEM, suggested that the internal stress is higher in the case of irradiated material than in the case of non-irradiated material, because of the very heterogeneous character of the plastic deformation. This analysis led to a coherent interpretation of the mechanical behavior of irradiated materials, in terms of deformation mechanisms. The mechanical behavior of irradiated materials was finally modeled by applying homogenization methods for heterogeneous materials. This model is able to reproduce adequately the mechanical behavior of the irradiated material, in agreement with the TEM observations. (author)

Determination of the neutral oxygen atom density in a plasma reactor loaded with metal samples

Science.gov (United States)

Mozetic, Miran; Cvelbar, Uros

2009-08-01

The density of neutral oxygen atoms was determined during processing of metal samples in a plasma reactor. The reactor was a Pyrex tube with an inner diameter of 11 cm and a length of 30 cm. Plasma was created by an inductively coupled radiofrequency generator operating at a frequency of 27.12 MHz and output power up to 500 W. The O density was measured at the edge of the glass tube with a copper fiber optics catalytic probe. The O atom density in the empty tube depended on pressure and was between 4 and 7 × 1021 m-3. The maximum O density was at a pressure of about 150 Pa, while the dissociation fraction of O2 molecules was maximal at the lowest pressure and decreased with increasing pressure. At about 300 Pa it dropped below 10%. The measurements were repeated in the chamber loaded with different metallic samples. In these cases, the density of oxygen atoms was lower than that in the empty chamber. The results were explained by a drain of O atoms caused by heterogeneous recombination on the samples.

Quantification of the distribution of hydrogen by nuclear microprobe at the Laboratory Pierre Sue in the width of zirconium alloy fuel clad of PWR reactors; Quantification de la repartition de l'hydrogene a la microsonde nucleaire du Laboratoire Pierre Sue dans l'epaisseur de tubes de gainage du combustible des REP en alliage de zirconium

Energy Technology Data Exchange (ETDEWEB)

Raepsaet, C. [CEA Saclay, Dept. de Recherche sur l' Etat Condense, les Atomes et les Molecules (DSM/DRECAM/LPS-CNRS) UMR9956, 91 - Gif sur Yvette (France); Bossis, Ph. [CEA Saclay, Dept. des Materiaux pour le Nucleaire (DEN/DANS/DMN/SEMULM2E), 91 - Gif-sur-Yvette (France); Hamon, D.; Bechade, J.L.; Brachet, J.C. [CEA Saclay, Dept. des Materiaux pour le Nucleaire (DEN/DANS/DMN/SRMALA2M), 91 - Gif-sur-Yvette (France)

2007-07-01

Among the analysis techniques by ions beams, the micro ERDA (Elastic Detection Analysis) is an interesting technique which allows the quantitative distribution of the hydrogen in materials. In particular, this analysis has been used for hydride zirconium alloys, with the nuclear microprobe of the Laboratory Pierre Sue. This probe allows the characterization of radioactive materials. The technique principles are recalled and then two examples are provided to illustrate the fuel clad behavior in PWR reactors. (A.L.B.)

Diagnostic study of the roughness surface effect of zirconium on the third-order nonlinear-optical properties of thin films based on zinc oxide nanomaterials

International Nuclear Information System (INIS)

Bahedi, K.; Addou, M.; El Jouad, M.; Sofiani, Z.; Alaoui Lamrani, M.; El Habbani, T.; Fellahi, N.; Bayoud, S.; Dghoughi, L.; Sahraoui, B.; Essaidi, Z.

2009-01-01

Zinc oxide (ZnO) and zirconium doped zinc oxide (ZnO:Zr) thin films were deposited by reactive chemical pulverization spray pyrolysis technique on heated glass substrates at 500 deg. C using zinc and zirconium chlorides as precursors. Effects of zirconium doping agent and surface roughness on the nonlinear optical properties were investigated in detail using atomic force microscopy (AFM) and third harmonic generation (THG) technique. The best value of nonlinear optical susceptibility χ (3) was obtained from the doped films with less roughness. A strong third order nonlinear optical susceptibility χ (3) = 20.12 x 10 -12 (esu) of the studied films was found for the 3% doped sample.

Supersonic pulsed free-jet of atoms and molecules of refractory metals: laser induced fluorescence spectroscopic studies on zirconium atoms and zirconium oxide molecules

International Nuclear Information System (INIS)

Nakhale, S.G.

2004-11-01

The experimental setup for generating supersonic pulsed free-jet containing atoms and molecules of refractory nature has been built. The technique of laser vaporization in conjunction with supersonic cooling is used to generate these species. The cooled atoms and molecules in supersonic free-jet are probed by laser induced fluorescence spectroscopy. In particular, the technique has been used to perform low-resolution laser induced fluorescence spectroscopy, limited by laser linewidth, on cold Zr atoms and ZrO molecules. The translational temperatures of ∼ 26.5 K and the rotational temperatures of ∼ 81 K have been achieved. It is possible to achieve the Doppler width of few tens of MHz allowing it to perform high-resolution spectroscopy on these atomic and molecular species. Also because of low rotational temperature of molecules the spectral congestion is greatly reduced. In general, this technique can be applied to perform spectroscopy on atoms and molecules of refractory nature. (author)

Applications for zirconium and columbium alloys

International Nuclear Information System (INIS)

Condliff, A.F.

1986-01-01

Currently, zirconium and columbium are used in a wide range of applications, overlapping only in the field of corrosion control. As a construction material, zirconium is primarily used by the nuclear power industry. The use of zirconium in the chemical processing industry (CPI) is, however, increasing steadily. Columbian alloys are primarily applied as superconducting alloys for research particle accelerators and fusion generators as well as in magnetic resonance imaging for medical diagnosis

Progress in zirconium resonance ionization spectroscopy

International Nuclear Information System (INIS)

Page, R.H.; Dropinski, S.C.; Worden, E.F.; Stockdale, J.A.D.

1993-01-01

The authors have examined the stepwise-resonant three-photon-ionization spectrum of neutral zirconium atoms using three separately-tunable pulsed visible dye lasers. The ground-level (first-step) transitions were chosen on the basis of demonstrated 91 Zr selectivity. Lifetimes of even-parity levels around 36,000 cm -1 , measured with the delayed-photoionization technique, range from 10 to 100 nsec. Direct ionization cross sections appear to be less than 10 -17 cm 2 ; newly-detected autoionizing levels give peak ionization cross sections (inferred from saturation fluences) up to 10 -15 cm 2 . Portions of Rydberg series converging to the 315 and 763 cm -1 levels of Zr + were identified. Clumps of autoionizing levels are thought to be due to Rydberg-valence mixing

Progress of the decommissioning process of Musashi Institute of Technology reactor (4)

International Nuclear Information System (INIS)

Uchiyama, Takafumi; Tanzawa, Tomio; Mitsuhashi, Ishi; Morishima, Kayoko; Matsumoto, Tetsuo

2012-01-01

The research reactor of Tokyo City University Atomic Energy Research Laboratory (Musashi Institute of Technology reactor) is zirconium-moderated water-cooled solid homogeneous type (TRIGA-II type), and its maximum heat output is 100 kW. It got into the first critical state in January 1963, and since then, it has mainly contributed to education and training for upgrading nuclear engineers, radioactivation analysis and reactor physics, and medical researches, as the joint usage research facilities across Japan. Then, after a long-term suspension, the university submitted the file in 2004 to the Ministry of Education, Culture, Sports, Science and Technology on the dismantling for the purpose of facility abolishment. Through the procedure of submitting a decommissioning plan, it was approved. Furthermore, in order to perform the function stop of the disposal facilities of liquid waste, application for change authorization for the decommissioning plan was submitted and approved. Regarding the progress of the decommissioning plan, the dismantling and removal of waste facilities for liquid waste and solid waste was carried out in FY2011 without any trouble. This paper explains this progress and future work plans. (A.O.)

Nuclear fuel fabrication in India

International Nuclear Information System (INIS)

Kondal Rao, N.

1975-01-01

The important role of a nuclear power programme in meeting the growing needs of power in India is explained. The successful installation of Tarapur Atomic Power Station and Rajasthan Atomic Power Station as well as the work at Madras Atomic Power Station are described. The development of the Atomic Fuels Division and the Nuclear Fuel Complex, Hyderabad which is mainly concerned with the fabrication of fuel elements and the reprocessing of fuels are explained. The N.F.C. essentially has the following constituent units : Zirconium Plant (ZP) comprising of Zirconium Oxide Plant, Zirconium Sponge Plant and Zirconium Fabrication Plant; Natural Uranium Oxide Plant (UOP); Ceramic Fuel Fabrication Plant (CFFP); Enriched Uranium Oxide Plant (EUOP); Enriched Fuel Fabrication Plant (EEFP) and Quality Control Laboratory for meeting the quality control requirements of all plants. The capacities of various plants at the NFC are mentioned. The work done on mixed oxide fuels and FBTR core with blanket assemblies, nickel and steel assemblies, thermal research reactor of 100 MW capacity, etc. are briefly mentioned. (K.B.)

Nuclear fuel fabrication in India

Energy Technology Data Exchange (ETDEWEB)

Kondal Rao, N

1975-01-01

The important role of a nuclear power program in meeting the growing needs of power in India is explained. The successful installation of Tarapur Atomic Power Station and Rajasthan Atomic Power Station as well as the work at Madras Atomic Power Station are described. The development of the Atomic Fuels Division and the Nuclear Fuel Complex, Hyderabad which is mainly concerned with the fabrication of fuel elements and the reprocessing of fuels are explained. The N.F.C. essentially has the following constituent units : Zirconium Plant (ZP) comprising of Zirconium Oxide Plant, Zirconium Sponge Plant and Zirconium Fabrication Plant; Natural Uranium Oxide Plant (UOP); Ceramic Fuel Fabrication Plant (CFFP); Enriched Uranium Oxide Plant (EUOP); Enriched Fuel Fabrication Plant (EEFP) and Quality Control Laboratory for meeting the quality control requirements of all plants. The capacities of various plants at the NFC are mentioned. The work done on mixed oxide fuels and FBTR core with blanket assemblies, nickel and steel assemblies, thermal research reactor of 100 MW capacity, etc. are briefly mentioned.

The development of zirconium alloy and its manufacturing

International Nuclear Information System (INIS)

Yuan Gaihuan; Yue Qiang

2015-01-01

Nuclear power which acts as one of low-carbon energy resources is the most realistic in large-scale application. It is also the preferred choice for many countries to develop energy resources and optimize its structure. Zirconium alloy is a key structural material for nuclear power plant fuel assemblies and cladding tubes of zirconium alloy are often referred as the first safeguard to nuclear power safety. With the development of nuclear power, three kinds of zirconium alloys Zr-Sn, Zr-Nb, Zr-Sn-Nb and with the representative products of Zr-4, M5, Zirlo respectively are developed and widely applied. Because of its severe operating environment and influence to nuclear safety, the requirements to zirconium alloys for physical and chemical properties, nuclear capability, tolerance and surface quality are very strict. The in-depth research and its manufacture capability become one of the main barriers for many countries who are developing the nuclear energy. In recent years, a stated-owned company, State Nuclear Bao Ti Zirconium Industry Company ('SNZ' for short) as well as National R and D Center for Nuclear Grade Zirconium material, is founded to meet the requirement of the rapid development of China's nuclear power industry. SNZ is dedicated for the fabrication and the research of nuclear grade zirconium products. After the successful completion of technology transfer of manufacturing for production chain and fully grasped of the manufacturing technology for the nuclear grade zirconium sponge through zirconium alloy tube, rod and strip products. National R and D Center for Nuclear Grade Zirconium material is cooperating with universities, nuclear energy research and design institutes and the owners of nuclear power plant to develop new zirconium alloy of self-owned brand. Through the selection of components, in-process testing and product inspection, four kinds of new zirconium alloys owns better performance than currently commercialized M5, Zirlo etc

A Novel Bis(phosphido)pyridine [PNP] 2− Pincer Ligand and Its Potassium and Bis(dimethylamido)zirconium(IV) Complexes

KAUST Repository

Winston, Matthew S.

2010-12-13

A novel PNP bis(secondary phosphine)pyridine pincer ligand, 2,6-bis(2-(phenylphosphino)phenyl)pyridine, has been prepared in high yield, and the properties of the doubly deprotonated form as a ligand in K 4(PNP)2(THF)6 and (PNP)Zr(NMe2) 2 have been investigated. The neutral PNP ligand has been isolated as a mixture of noninterconverting diastereomers, due to the presence of two chirogenic phosphorus atoms of the secondary phopshines, but coordination of the dianionic form to potassium and zirconium allows for isolation of a single diastereomer in near-quantitative yield. The structure of a bis(dimethylamido) zirconium(IV) derivative of the bis(phosphido)pyridine ligand and DFT calculations suggest that the phosphides do not π-bond to early transition metals, likely due to geometric strain and possibly orbital size mismatch between phosphorus and zirconium. As a result, the soft phosphides are prone to formation of insoluble oligomers with substantial bridging of the phosphido lone pairs to other zirconium centers. © 2010 American Chemical Society.

A Novel Bis(phosphido)pyridine [PNP] 2− Pincer Ligand and Its Potassium and Bis(dimethylamido)zirconium(IV) Complexes

KAUST Repository

Winston, Matthew S.; Bercaw, John E.

2010-01-01

A novel PNP bis(secondary phosphine)pyridine pincer ligand, 2,6-bis(2-(phenylphosphino)phenyl)pyridine, has been prepared in high yield, and the properties of the doubly deprotonated form as a ligand in K 4(PNP)2(THF)6 and (PNP)Zr(NMe2) 2 have been investigated. The neutral PNP ligand has been isolated as a mixture of noninterconverting diastereomers, due to the presence of two chirogenic phosphorus atoms of the secondary phopshines, but coordination of the dianionic form to potassium and zirconium allows for isolation of a single diastereomer in near-quantitative yield. The structure of a bis(dimethylamido) zirconium(IV) derivative of the bis(phosphido)pyridine ligand and DFT calculations suggest that the phosphides do not π-bond to early transition metals, likely due to geometric strain and possibly orbital size mismatch between phosphorus and zirconium. As a result, the soft phosphides are prone to formation of insoluble oligomers with substantial bridging of the phosphido lone pairs to other zirconium centers. © 2010 American Chemical Society.

International Atomic Energy Agency (IAEA) Activity on Technical Influence of High Burnup UOX and MOX Water Reactor Fuel on Spent Fuel Management

International Nuclear Information System (INIS)

Lovasic, Z.; Einziger, R.

2009-01-01

This paper briefly reviews the results of the International Atomic Energy Agency (IAEA) project investigating the influence of high burnup and mixed-oxide (MOX) fuels, from water power reactors, on spent fuel management. These data will provide information on the impacts, regarding spent fuel management, for those countries operating light-water reactors (LWR)s and heavy-water reactors (HWR)s with zirconium alloy-clad uranium dioxide (UOX) fuels, that are considering the use of higher burnup UOX or the introduction of reprocessing and MOX fuels. The mechanical designs of lower burnup UOX and higher burnup UOX or MOX fuel are very similar, but some of the properties (e.g., higher fuel rod internal pressures; higher decay heat; higher specific activity; and degraded cladding mechanical properties of higher burnup UOX and MOX spent fuels) may potentially significantly affect the behavior of the fuel after irradiation. These properties are reviewed. The effects of these property changes on wet and dry storage, transportation, reprocessing, re-fabrication of fuel, and final disposal were evaluated, based on regulatory, safety, and operational considerations. Political and strategic considerations were not taken into account since relative importance of technical, economic and strategic considerations vary from country to country. There will also be an impact of these fuels on issues like non-proliferation, safeguards, and sustainability, but because of the complexity of factors affecting those issues, they are only briefly discussed. Data gaps were also identified during this investigation. The pros and cons of using high burnup UOX or MOX, for each applicable issue in each stage of the back end of the fuel cycle, were evaluated and are discussed.. Although, in theory, higher burnup fuel and MOX fuels mean a smaller quantity of spent fuel, the potential need for some changes in design of spent fuel storage, transportation, handling, reprocessing, re-fabrication, and

Irradiation growth in zirconium alloys: a review

International Nuclear Information System (INIS)

Fidleris, V.

1980-09-01

The change in shape during irradiation without external stress, irradiation growth, was first discovered in uranium and later in graphite, zirconium and other core materials which exhibit anisotropic physical properties. The direction of maximum growth of metals invariably corresponds with the direction of minimum thermal expansion. In polycrystalline zirconium alloys growth is positive in the direction of maximum deformation during fabrication and in other directions it can be either positive or negative depending on the preferred orientation of grains (crystallographic texture). Growth increases gradually with temperature between 300 K and 620 K and rapidly with fluence up to about 1 x 10 25 n.m. -2 (Eμ1 MeV). At higher fluences the growth appears to saturate in annealed materials and reach a steady rate approximately proportional to dislocation density in cold-worked materials. Above 600 K both annealed and cold-worked materials have similar steady growth rates. Irradiation growth is caused by the segregation to different sinks of the vacancies and interstitials generated by irradiation, but the dominant types of sinks for each type of point defect and the mode of transport of the point defects to sinks cannot therefore be predicted theoretically. For the purpose of designing reactor core components empirical equations have been derived that can satisfactorily predict the steady state growth behaviour from texture and microstructure. (auth)

Atomistic studies of cation transport in tetragonal ZrO2 during zirconium corrosion

International Nuclear Information System (INIS)

Bai, Xian-Ming; Zhang, Yongfeng; Tonks, Michael R.

2015-01-01

Zirconium alloys are the major fuel cladding materials in current reactors. The water-side corrosion is a significant degradation mechanism of these alloys. During corrosion, the transport of oxidizing species in zirconium dioxide (ZrO 2 ) determines the corrosion kinetics. Previously, it has been argued that the outward diffusion of cations is important for forming protective oxides. In this work, the migration of Zr defects in tetragonal ZrO 2 is studied with temperature accelerated dynamics and molecular dynamics simulations. The results show that Zr interstitials have anisotropic diffusion and migrate preferentially along the [001] or c direction in tetragonal ZrO 2 . The compressive stresses can increase the Zr interstitial migration barrier significantly. The migration of Zr interstitials at a grain boundary is much slower than in a bulk oxide. The implications of these atomistic simulation results in the Zr corrosion are discussed. (authors)

Oxidized zirconium on ceramic; Catastrophic coupling.

Science.gov (United States)

Ozden, V E; Saglam, N; Dikmen, G; Tozun, I R

2017-02-01

Oxidized zirconium (Oxinium™; Smith & Nephew, Memphis, TN, USA) articulated with polyethylene in total hip arthroplasty (THA) appeared to have the potential to reduce wear dramatically. The thermally oxidized metal zirconium surface is transformed into ceramic-like hard surface that is resistant to abrasion. The exposure of soft zirconium metal under hard coverage surface after the damage of oxidized zirconium femoral head has been described. It occurred following joint dislocation or in situ succeeding disengagement of polyethylene liner. We reported three cases of misuse of Oxinium™ (Smith & Nephew, Memphis, TN, USA) heads. These three cases resulted in catastrophic in situ wear and inevitable failure although there was no advice, indication or recommendation for this use from the manufacturer. Copyright © 2016 Elsevier Masson SAS. All rights reserved.

Determination of hafnium by atomic absorption spectrometry

International Nuclear Information System (INIS)

Yoshida, Isao; Kobayashi, Hiroshi; Ueno, Keihei.

1977-01-01

Optimum conditions for atomic absorption spectrophotometric determination of hafnium were investigated by use of a Jarrel-Ash AA-1 instrument which was equipped with a premixed gas burner slotted 50 mm in length and 0.4 mm in width. Absorption of hafnium, which was atomized in an nitrous oxide-acetylene flame, was measured on a resonance line at 307.29 nm. The absorption due to hafnium was enhanced in the presence of ammonium fluoride and iron(III) ion, as shown in Figs. 2 and 3, depending on their concentration. The highest absorption was attained by the addition of (0.15 -- 0.3)M ammonium fluoride, 0.07 M of iron(III) ion and 0.05 M of hydrochloric acid. An excess of the additives decreased the absorption. The presence of zirconium, which caused a significant interference in the ordinary analytical methods, did not affect the absorption due to hafnium, if the zirconium concentration is less than 0.2 M. A standard procedure was proposed; A sample containing a few mg of hafnium was dissolved in a 25-ml volumetric flask, and ammonium fluoride, ferric nitrate and hydrochloric acid were added so that the final concentrations were 0.3, 0.07 and 0.05 M, respectively. Atomic absorption was measured on the aqueous solution in a nitrous oxide-acetylene flame and the hafnium content was calculated from the absorbance. Sensitivity was as high as 12.5 μg of Hf/ml/l% absorption. The present method is especially recommendable to the direct determination of hafnium in samples containing zirconium. (auth.)

Electrochemical-metallothermic reduction of zirconium in molten salt solutions

International Nuclear Information System (INIS)

McLaughlin, D.F.; Talko, F.

1990-01-01

This patent describes a method for separating hafnium from zirconium of the type wherein a feed containing zirconium and hafnium chlorides is prepared from zirconium-hafnium chloride and the feed is introduced into a distillation column, which distillation column has a reboiler connected at the bottom and a reflux condenser connected at the top and wherein a hafnium chloride enriched stream is taken from the top of the column and a zirconium enriched chloride stream is taken from the bottom of the column. It comprises: reducing the zirconium enriched chloride stream taken from the distillation column to metal by electrochemically reducing an alkaline earth metal in a molten salt bath with the molten salt in the molten salt bath consisting essentially of a mixture of at least one alkali metal chloride and at least one alkaline earth metal chloride and zirconium chloride, with the reduced alkaline earth metal reacting with the zirconium chloride to produce zirconium metal and alkaline earth metal chloride

Dissolution of anodic zirconium dioxide films in aqueous media

International Nuclear Information System (INIS)

Merati, A.; Cox, B.

1999-01-01

Zirconium with a low thermal neutron cross section, good corrosion resistance in high-temperature water, and high thermal conductivity is an ideal material for nuclear reactors. Its good resistance to water and steam at reactor temperatures is of the greatest interest to nuclear fuel designers. Dissolution of zirconium dioxide (ZrO 2 ) films in aggressive media was investigated. The extent of uniform and localized dissolution was measured by ultraviolet-visible (UV-VIS) spectrometry and an alternating current (AC) impedance test, respectively. Scanning electron microscopy (SEM) showed the extent of dissolution of ZrO 2 was a function only of the fluoride ion content and pH of the medium. Cathodic polarization was used to identify the preferred sites for localized dissolution of the oxide film. In 0.1 M potassium bifluoride (KHF 2 ), both uniform thinning and local breakdown of the oxide were observed. Within the limits of the investigating techniques, no evidence of dissolution was observed in the other solutions tested: 0.5 M sulfuric acid (H 2 SO 4 ). 1.0 M nitric acid (HNO 3 ), 5 M hydrochloric acid (HCl), or 0.1 M potassium fluoride (KF). In areas around iron-containing particles, fine cracks in the anodic oxide at prior metal grain boundaries and arrays of cracks in the oxide associated with residual scratches from the initial specimen preparation were the preferred spots for localized dissolution of the oxide film. Iron precipitates immediately below the surface of the oxide layer increased the local electrical conductivity. Enrichment of iron in the oxide matrix around these precipitates during the anodization process appeared to cause prospective spots, acting as anodic sites for pH formation

Effects of operating conditions on molten-salt electrorefining for zirconium recovery from irradiated Zircaloy-4 cladding of pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Park, Jaeyeong, E-mail: d486916@snu.ac.kr [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of); Choi, Sungyeol [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Sohn, Sungjune [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of); Kim, Kwang-Rag [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Hwang, Il Soon [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of)

2014-08-15

Highlights: • Computational simulation on electrorefining of irradiated Zircaloy-4 cladding. • Composition of irradiated Zircaloy-4 cladding of pressurized water reactor. • Redox behavior of elements in irradiated Zircaloy cladding during electrorefining. • Effect of electrorefining operating conditions on decontamination factor. - Abstract: To reduce the final waste volume from used nuclear fuel assembly, it is significant to decontaminate irradiated cladding. Electrorefining in high temperature molten salt could be one of volume decontamination processes for the cladding. This study examines the effect of operating conditions on decontamination factor in electrorefining of irradiated Zircaloy-4 cladding of pressurized water reactor. One-dimensional time-dependent electrochemical reaction code, REFIN, was utilized for simulating irradiated cladding electrorefining. Composition of irradiated Zircaloy was estimated based on ORIGEN-2 and other literatures. Co and U were considered in electrorefining simulation with major elements of Zircaloy-4 to represent activation products and actinides penetrating into the cladding respectively. Total 240 cases of electrorefining are simulated including 8 diffusion boundary layer thicknesses, 10 concentrations of contaminated molten salt and 3 termination conditions. Decontamination factors for each case were evaluated and it is revealed that the radioactivity of Co-60 in recovered zirconium on cathode could decrease below the clearance level when initial concentration of chlorides except ZrCl{sub 4} is lower than 1 × 10{sup −11} weight fraction if electrorefining is finished before anode potential reaches −1.8 V (vs. Cl{sub 2}/Cl{sup −})

Experimental studies of U-Pu-Zr fast reactor fuel pins in EBR-II [Experimental Breeder Reactor

International Nuclear Information System (INIS)

Pahl, R.G.; Porter, D.L.; Lahm, C.E.; Hofman, G.L.

1988-01-01

The Integral Fast Reactor (IFR) is a generic reactor concept under development by Argonne National Laboratory. Much of the technology for the IFR is being demonstrated at the Experimental Breeder Reactor II (EBR-II) on the Department of Energy site near Idaho Falls, Idaho. The IFR concept relies on four technical features to achieve breakthroughs in nuclear power economics and safety: (1) a pool-type reactor configuration, (2) liquid sodium cooling, (3) metallic fuel, and (4) an integral fuel cycle with on-site reprocessing. The purpose of this paper will be to summarize our latest results of irradiation testing uranium-plutonium-zirconium (U-Pu-Zr) fuel in the EBR-II. 10 refs., 13 figs., 2 tabs

The near boiling reactor: design of a small nuclear reactor for extending the operational envelope of the Victoria Class Submarine

International Nuclear Information System (INIS)

Cole, C.; Bonin, H.

2005-01-01

A small, inherently safe nuclear reactor that will provide enough power to maintain the hotel load of the Victoria Class Submarine and extend her operational envelope, has been conceptually designed. The final reactor concept, named the Near Boiling (NB) Reactor, employs TRISO fuel particles in Zirconium cladded fuel rods. The reactor is light water moderated and cooled. The core life is specifically designed to coincide with the refit cycle of the Victoria Class Submarine. The reactor employs a simple and reliable control and shut down system that requires little intervention on the part of the submarine's crew. Also, a kinetic model is developed that demonstrates the inherent safety features of the reactor during several accident scenarios. (author)

The near boiling reactor: design of a small nuclear reactor for extending the operational envelope of the Victoria Class Submarine

Energy Technology Data Exchange (ETDEWEB)

Cole, C.; Bonin, H. [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)]. E-mail: chris.cole@rmc.ca; bonin-h@rmc.ca

2005-07-01

A small, inherently safe nuclear reactor that will provide enough power to maintain the hotel load of the Victoria Class Submarine and extend her operational envelope, has been conceptually designed. The final reactor concept, named the Near Boiling (NB) Reactor, employs TRISO fuel particles in Zirconium cladded fuel rods. The reactor is light water moderated and cooled. The core life is specifically designed to coincide with the refit cycle of the Victoria Class Submarine. The reactor employs a simple and reliable control and shut down system that requires little intervention on the part of the submarine's crew. Also, a kinetic model is developed that demonstrates the inherent safety features of the reactor during several accident scenarios. (author)

Process and equipement for zone heat treatment of zirconium alloys tubes

International Nuclear Information System (INIS)

Kiesler, A.J.; Frischmann, P.G.; Rockwood, A.C.

1977-01-01

Process for the thermal treatment of an area of a long zirconium alloy part in order to increase its corrosion resistance in the cooling conditions of boiling water reactor, in which the part is moved lengthwise through a succession of critical maximum temperature areas, during a critical time and is subjected to a temperature reduction at critical rate, so that each successive portion reaches a maximum temperature between 825 0 C, and directing water at a temperature around 60 to 80 0 C as jets in the cooling area [fr

Operation and maintenance experience at the General Atomic Company's TRIGA reactor facility at San Diego, California

International Nuclear Information System (INIS)

Whittemore, W.L.; Stout, W.A.; Shoptaugh, J.R.; Chesworth, R.H.

1982-01-01

Since the startup of the original 250 kW TRIGA Mark I reactor in 1958, General Atomic Company has accumulated nearly 24 years of operation and maintenance experience with this type of reactor. In addition to the nearly 24 years of experience gained on the Mark I, GA has operated the 1.5 MW Advanced Prototype Test Reactor (Mark F) for 22 years and operated a 2 MW below-ground TRIGA Mark III for five years. Information obtained from normal and abnormal operation are presented. (author)

Problems of zirconium metal production in Czechoslovakia

International Nuclear Information System (INIS)

Vareka, J.; Vaclavik, E.

1975-01-01

The problems are summed up of the production and quality control of zirconium sponge. A survey is given of industrial applications of zirconium in form of pure metal or alloys in nuclear power production, ferrous and non-ferrous metallurgy, chemical engineering and electrical engineering. A survey is also presented of the manufacture of zirconium metal in advanced capitalist countries. (J.B.)

Intercomparison of techniques for inspection and diagnostics of heavy water reactor pressure tubes. Determination of hydrogen concentration and blister characterization

International Nuclear Information System (INIS)

2009-03-01

Heavy water reactors (HWRs) comprise significant numbers of today's operating nuclear power plants, and more are under construction. Efficient and accurate inspection and diagnostic techniques for various reactor components and systems, especially pressure tubes, are an important factor in ensuring reliable and safe plant operation. To foster international collaboration in the efficient and safe use of nuclear power, the IAEA conducted a Coordinated Research Project (CRP) on Intercomparison of Techniques for HWR Pressure Tube Inspection and Diagnostics. This CRP was carried out within the framework of the IAEA's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). The TWG-HWR is a group of experts nominated by their governments and designated by the IAEA to provide advice and to support implementation of IAEA's project on advanced technologies for HWRs. The objective of the CRP was to compare non-destructive inspection and diagnostic techniques, in use and being developed, for structural integrity assessment of HWR pressure tubes. During the first phase of this CRP participants investigated the capability of different techniques to detect and characterize flaws. During the second phase participants collaborated to detect and characterize hydride blisters and to determine the hydrogen concentration in zirconium alloys. The intention was to identify the most effective pressure tube inspection and diagnostic methods and to identify further development needs. The organizations which participated in phase 2 of this CRP are: - Comision Nacional de Energia Atomica (CNEA), Argentina; - Atomic Energy of Canada Ltd. (AECL), Chalk River Laboratories (CRL), Canada; - Bhabha Atomic Research Centre (BARC), India; - Korea Atomic Energy Research Institute (KAERI), Republic of Korea; - National Institute for Research and Development for Technical Physics (NIRDTP), Romania; - Nuclear Non-Destructive Testing Research and Services (NNDT), Romania. IAEA-TECDOC-1499

Study of diffusion processes in the oxide layer of zirconium alloys

Directory of Open Access Journals (Sweden)

Sialini P.

2016-03-01

Full Text Available In the active zone of a nuclear reactor where zirconium alloys are used as a coating material, this material is subject to various harmful impacts. During water decomposition reactions, hydrogen and oxygen are evolved that may diffuse through the oxidic layer either through zirconium dioxide (ZrO2 crystals or along ZrO2 grains. The diffusion mechanism can be studied using the Ion Beam Analysis (IBA method where nuclear reaction 18O(p,α15N is used. A tube made of zirconium alloy E110 (with 1 wt. % of Nb was used for making samples that were pre-exposed in UJP PRAHA a.s. and subsequently exposed to isotopically cleansed environment of H2 18O medium in an autoclave. The samples were analysed with gravimetric methods and IBA methods performed at the electrostatic particle accelerator Tandetron 4130 MC in the Nucler Physics Institute of the CAS, Řež. With IBA methods, the overall thicknesses of corrosion layers on the samples, element composition of the alloy and distribution of oxygen isotope 18O in the corrosion layer and its penetration in the alloy were identified. The retrieved data shows at the oxygen diffusion along ZrO2 grains because there are two peaks of 18O isotope concentrations in the corrosion layer. These peaks occur at the environment-oxide and oxide-metal interface. The element analysis identified the presence of undesirable hafnium.

Expanded heat treatment to form residual compressive hoop stress on inner surface of zirconium alloy tubing

International Nuclear Information System (INIS)

Megata, Masao

1997-01-01

A specific heat treatment process that introduces hoop stress has been developed. This technique can produce zirconium alloy tubing with a residual compressive hoop stress near the inner surface by taking advantage of the mechanical anisotropy in hexagonal close-packed zirconium crystal. Since a crystal having its basal pole parallel to the tangential direction of the tubing is easier to exhibit plastic elongation under the hoop stress than that having its basal pole parallel to the radial direction, the plastic and elastic elongation can coexist under a certain set of temperature and hoop stress conditions. The mechanical anisotropy plays a role to extend the coexistent stress range. Thus, residual compressive hoop stress is formed at the inner surface where more plastic elongation occurs during the heat treatment. This process is referred to as expanded heat treatment. Since this is a fundamental crystallographic principle, it has various applications. The application to improve PCI/SCC (pellet cladding interaction/stress corrosion cracking) properties of water reactor fuel cladding is promising. Excellent results were obtained with laboratory-scale heat treatment and an out-reactor iodine SCC test. These results included an extension of the time to SCC failure. (author)

Conversion, core redesign and upgrade of the Rhode Island Atomic Energy Commission Reactor

International Nuclear Information System (INIS)

DiMeglio, A.F.

1987-01-01

The 2 MW Rhode Island Atomic Energy Commission reactor is required to convert from the use of High Enriched Uranium (HEU) fuel to the use of Low Enriched Uranium (LEU) fuel using a standard LEU fuel plate which is thinner and contains more Uranium-235 than the current HEU plate. These differences, coupled with the fact that the conversion should be accomplished without serious degradation of reactor characteristics and capability, has resulted in core design studies and thermal hydraulic studies not only at the current 2 MW but also at the maximum power level of the reactor, 5 MW. In addition, during the course of its 23 years of operation, it has become clear that the main uses of the reactor are neutron scattering and neutron activation analysis. The requirement to convert to LEU presents an opportunity during the conversion to optimize the core for the utilization and to restudy the thermal hydraulics using modern techniques. This paper will present the preliminary conclusions of both aspects. (Author)

Phase Transformations in a Uranium-Zirconium Alloy containing 2 weight per cent Zirconium

Energy Technology Data Exchange (ETDEWEB)

Lagerberg, G

1961-04-15

The phase transformations in a uranium-zirconium alloy containing 2 weight percent zirconium have been examined metallographically after heat treatments involving isothermal transformation of y and cooling from the -y-range at different rates. Transformations on heating and cooling have also been studied in uranium-zirconium alloys with 0.5, 2 and 5 weight per cent zirconium by means of differential thermal analysis. The results are compatible with the phase diagram given by Howlett and Knapton. On quenching from the {gamma}-range the {gamma} phase transforms martensitically to supersaturated a the M{sub S} temperature being about 490 C. During isothermal transformation of {gamma} in the temperature range 735 to 700 C {beta}-phase is precipitated as Widmanstaetten plates and the equilibrium structure consists of {beta} and {gamma}{sub 1}. Below 700 C {gamma} transforms completely to Widmanstaetten plates which consist of {beta} above 660 C and of a at lower temperatures. Secondary phases, {gamma}{sub 2} above 610 C and {delta} below this temperature, are precipitated from the initially supersaturated Widmanstaetten plates during the isothermal treatments. At and slightly below 700 C the cooperative growth of |3 and {gamma}{sub 2} is observed. The results of isothermal transformation are summarized in a TTTdiagram.

Diagnostic study of the roughness surface effect of zirconium on the third-order nonlinear-optical properties of thin films based on zinc oxide nanomaterials

Energy Technology Data Exchange (ETDEWEB)

Bahedi, K., E-mail: bahedikhadija@yahoo.com [Laboratoire Optoelectronique et Physico-chimie des Materiaux Universite Ibn Tofail, Faculte des Sciences BP 133 Kenitra 14000, Maroc (Morocco); Addou, M.; El Jouad, M.; Sofiani, Z.; Alaoui Lamrani, M.; El Habbani, T.; Fellahi, N.; Bayoud, S.; Dghoughi, L. [Laboratoire Optoelectronique et Physico-chimie des Materiaux Universite Ibn Tofail, Faculte des Sciences BP 133 Kenitra 14000, Maroc (Morocco); Sahraoui, B.; Essaidi, Z. [Laboratoire POMA, UMR CNRS 6136, Universite d' Angers 2, Bd Lavoisier, 49045 France (France)

2009-02-01

Zinc oxide (ZnO) and zirconium doped zinc oxide (ZnO:Zr) thin films were deposited by reactive chemical pulverization spray pyrolysis technique on heated glass substrates at 500 deg. C using zinc and zirconium chlorides as precursors. Effects of zirconium doping agent and surface roughness on the nonlinear optical properties were investigated in detail using atomic force microscopy (AFM) and third harmonic generation (THG) technique. The best value of nonlinear optical susceptibility {chi}{sup (3)} was obtained from the doped films with less roughness. A strong third order nonlinear optical susceptibility {chi}{sup (3)} = 20.12 x 10{sup -12} (esu) of the studied films was found for the 3% doped sample.

In-reactor behaviour of centrifugally atomized U3Si dispersion fuel irradiated at high temperature in HANARO

International Nuclear Information System (INIS)

Kim, Ki Hwan; Park, Jong Man; Yoo, Byeong Ok; Park, Dae Kyu; Lee, Choong Sung; Kim, Chang Kyu

2002-01-01

The irradiation test on full-size U 3 Si dispersion fuel elements, prepared by centrifugal atomization and conventional comminution method, has been performed up to about 77 at.% U-235 in maximum burn-up at CT hole position having the highest power condition in the HANARO reactor, in order to examine the irradiation performance of the atomized U 3 Si for the driver fuels of HANARO. The in-reactor interaction of the atomized U 3 Si dispersion fuel meats is generally assumed to be acceptable with the range of 5-15 μm in average thickness. The atomized spherical particles have more uniform and thinner reaction layer than the comminuted irregular particles. The U 3 Si particles have relatively fine and uniform size distribution of fission gas bubbles, irrespective of the powdering method. The bubble population in the atomized particles appears to be finer and more homogeneous with the characteristics of narrower bubble size distribution than that of the comminuted fuel. The atomized U 3 Si dispersion fuel elements exhibit sound swelling behaviours of 5 % in ΔV/V m even at ∼77 at.% U-235 burn-up, which meets with the safety criterion of the fuel rod, 20vol.% for HANARO. The atomized U3Si dispersion fuel elements show smaller swelling than the comminuted fuel elements

Role of plasma enhanced atomic layer deposition reactor wall conditions on radical and ion substrate fluxes

Energy Technology Data Exchange (ETDEWEB)

Sowa, Mark J., E-mail: msowa@ultratech.com [Ultratech/Cambridge NanoTech, 130 Turner Street, Building 2, Waltham, Massachusetts 02453 (United States)

2014-01-15

Chamber wall conditions, such as wall temperature and film deposits, have long been known to influence plasma source performance on thin film processing equipment. Plasma physical characteristics depend on conductive/insulating properties of chamber walls. Radical fluxes depend on plasma characteristics as well as wall recombination rates, which can be wall material and temperature dependent. Variations in substrate delivery of plasma generated species (radicals, ions, etc.) impact the resulting etch or deposition process resulting in process drift. Plasma enhanced atomic layer deposition is known to depend strongly on substrate radical flux, but film properties can be influenced by other plasma generated phenomena, such as ion bombardment. In this paper, the chamber wall conditions on a plasma enhanced atomic layer deposition process are investigated. The downstream oxygen radical and ion fluxes from an inductively coupled plasma source are indirectly monitored in temperature controlled (25–190 °C) stainless steel and quartz reactors over a range of oxygen flow rates. Etch rates of a photoresist coated quartz crystal microbalance are used to study the oxygen radical flux dependence on reactor characteristics. Plasma density estimates from Langmuir probe ion saturation current measurements are used to study the ion flux dependence on reactor characteristics. Reactor temperature was not found to impact radical and ion fluxes substantially. Radical and ion fluxes were higher for quartz walls compared to stainless steel walls over all oxygen flow rates considered. The radical flux to ion flux ratio is likely to be a critical parameter for the deposition of consistent film properties. Reactor wall material, gas flow rate/pressure, and distance from the plasma source all impact the radical to ion flux ratio. These results indicate maintaining chamber wall conditions will be important for delivering consistent results from plasma enhanced atomic layer deposition

Dynamic Modeling for the Design and Cyclic Operation of an Atomic Layer Deposition (ALD Reactor

Directory of Open Access Journals (Sweden)

Curtisha D. Travis

2013-08-01

Full Text Available A laboratory-scale atomic layer deposition (ALD reactor system model is derived for alumina deposition using trimethylaluminum and water as precursors. Model components describing the precursor thermophysical properties, reactor-scale gas-phase dynamics and surface reaction kinetics derived from absolute reaction rate theory are integrated to simulate the complete reactor system. Limit-cycle solutions defining continuous cyclic ALD reactor operation are computed with a fixed point algorithm based on collocation discretization in time, resulting in an unambiguous definition of film growth-per-cycle (gpc. A key finding of this study is that unintended chemical vapor deposition conditions can mask regions of operation that would otherwise correspond to ideal saturating ALD operation. The use of the simulator for assisting in process design decisions is presented.

The fluorimetric titration of zirconium in the ppm-range

International Nuclear Information System (INIS)

Linden, W.E. von der; Boef, G. den; Ozinga, W.

1976-01-01

A fluorimetric titration of zirconium(IV) with EDTA is proposed. The fluorescence intensity of the zirconium-morin complex is used to indicate the end-point. More than twenty other cations were investigated and it was found that they did not interfere, neither did common anions. Mercury(II) can only be tolerated in amount not exceeding that of zirconium. Bismuth(III) interferes and hafnium(IV0 is titrated together with zirconium. The relative standard deviation of the titration of 10ml of a solution containing 1 ppm of zirconium does not exceed 1.5%

Zirconium behaviour during electrorefining of actinide-zirconium alloy in molten LiCl-KCl on aluminium cathodes

Energy Technology Data Exchange (ETDEWEB)

Meier, R. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, Karlsruhe 76125 (Germany); Heidelberg University, Institute of Physical Chemistry, Im Neuenheimer Feld 253, Heidelberg 69120 (Germany); Souček, P., E-mail: Pavel.Soucek@ec.europa.eu [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, Karlsruhe 76125 (Germany); Malmbeck, R.; Krachler, M.; Rodrigues, A.; Claux, B.; Glatz, J.-P. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, Karlsruhe 76125 (Germany); Fanghänel, Th. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, Karlsruhe 76125 (Germany); Heidelberg University, Institute of Physical Chemistry, Im Neuenheimer Feld 253, Heidelberg 69120 (Germany)

2016-04-15

A pyrochemical electrorefining process for the recovery of actinides from metallic nuclear fuel based on actinide-zirconium alloys (An–Zr) in a molten salt is being investigated. In this process actinides are group-selectively recovered on solid aluminium cathodes as An–Al alloys using a LiCl–KCl eutectic melt at a temperature of 450 °C. In the present study the electrochemical behaviour of zirconium during electrorefining was investigated. The maximum amount of actinides that can be oxidised without anodic co-dissolution of zirconium was determined at a selected constant cathodic current density. The experiment consisted of three steps to assess the different stages of the electrorefining process, each of which employing a fresh aluminium cathode. The results indicate that almost a complete dissolution of the actinides without co-dissolution of zirconium is possible under the applied experimental conditions. - Highlights: • Recovery of actinides was shown by electrorefining of U/Pu–Zr alloys in LiCl–KCl. • Constant current density of 20 mA/cm{sup 2} is applied. • Most of the actinides were dissolved avoiding zirconium co-dissolution. • Deterioration of the deposit quality by a small amount of co-deposited Zr is not observed.

SEPARATING HAFNIUM FROM ZIRCONIUM

Science.gov (United States)

Lister, B.A.J.; Duncan, J.F.

1956-08-21

A dilute aqueous solution of zirconyl chloride which is 1N to 2N in HCl is passed through a column of a cation exchange resin in acid form thereby absorbing both zirconium and associated hafnium impurity in the mesin. The cation exchange material with the absorbate is then eluted with aqueous sulfuric acid of a O.8N to 1.2N strength. The first portion of the eluate contains the zirconium substantially free of hafnium.

High temperature brazing of reactor materials

International Nuclear Information System (INIS)

Orlov, A.V.; Nechaev, V.A.; Rybkin, B.V.; Ponimash, I.D.

1990-01-01

Application of high-temperature brazing for joining products of such materials as molybdenum, tungsten, zirconium, beryllium, magnesium, nickel and aluminium alloys, graphite ceramics etc. is described. Brazing materials composition and brazed joints properties are presented. A satisfactory strength of brazed joints is detected under reactor operation temperatures and coolant and irradiation effect

Atomic layer deposition of zirconium dioxide from zirconium tetrachloride and ozone

Energy Technology Data Exchange (ETDEWEB)

Kukli, Kaupo, E-mail: kaupo.kukli@helsinki.fi [Department of Chemistry, University of Helsinki, P.O. Box 55, FI-00014 Helsinki (Finland); Kemell, Marianna; Köykkä, Joel [Department of Chemistry, University of Helsinki, P.O. Box 55, FI-00014 Helsinki (Finland); Mizohata, Kenichiro [Accelerator Laboratory, Department of Physics, University of Helsinki, P.O. Box 43, FI-00014 Helsinki (Finland); Vehkamäki, Marko; Ritala, Mikko; Leskelä, Markku [Department of Chemistry, University of Helsinki, P.O. Box 55, FI-00014 Helsinki (Finland)

2015-08-31

ZrO{sub 2} films were grown by atomic layer deposition using ZrCl{sub 4} and O{sub 3} as precursors. The films were grown on silicon substrates in the temperature range of 220–500 °C. The ALD rate was monotonously decreasing from 0.085 to 0.060 nm/cycle in this temperature range towards the highest temperatures studied. The content of chlorine in the films did not exceed 0.2 at.% as measured by elastic recoil detection analysis. The content of hydrogen was 0.30 and 0.14 at.% in the films grown at 300 and 400 °C, respectively. Structural studies revealed the films consisting of mixtures of stable monoclinic and metastable tetragonal/cubic polymorphs of ZrO{sub 2}, and dominantly metastable phases of ZrO{sub 2} below and above 300 °C, respectively. Permittivity of dielectric layers in Al/Ti/ZrO{sub 2}/(TiN/)Si capacitors with 15–40 nm thick ZrO{sub 2} ranged between 12 and 25 at 100 kHz and the dielectric breakdown fields were in the range of 1.5–3.0 MV/cm. - Highlights: • ZrO{sub 2} thin films were grown by atomic layer deposition from ZrCl{sub 4} and O{sub 3}. • Relatively high substrate temperatures promoted growth of metastable ZrO{sub 2} phases. • ZrO{sub 2} films exhibited electric properties characteristic of dielectric metal oxides. • ZrO{sub 2} grown in hydrogen- and carbon free process contained low amounts of impurities.

Spectrophotometric titration of zirconium in siliceous materials

International Nuclear Information System (INIS)

Sugawara, K.F.; Su, Y.-S.; Strzegowski, W.R.

1978-01-01

An accurate and selective complexometric titration procedure based upon a spectrophotometrically detected end-point has been developed for the determination of zirconium in glasses, glass-ceramics and refractories. A p-bromomandelic acid separation step for zirconium imparts excellent selectivity to the procedure. The method is particularly important for the 1 to 5% concentration range where a simple, accurate and selective method for the determination of zirconium has been lacking. (author)

Room-temperature atomic layer deposition of ZrO{sub 2} using tetrakis(ethylmethylamino)zirconium and plasma-excited humidified argon

Energy Technology Data Exchange (ETDEWEB)

Kanomata, K. [Graduate School of Science and Engineering, Yamagata University, 4-3-16 Jonan, Yonezawa 992-8510 (Japan); Japan Society for the Promotion of Science, 5-3-1 Kojimachi, Chiyoda-ku, Tokyo 102-0083 (Japan); Tokoro, K.; Imai, T.; Pansila, P.; Miura, M.; Ahmmad, B.; Kubota, S.; Hirahara, K. [Graduate School of Science and Engineering, Yamagata University, 4-3-16 Jonan, Yonezawa 992-8510 (Japan); Hirose, F., E-mail: fhirose@yz.yamagata-u.ac.jp [Graduate School of Science and Engineering, Yamagata University, 4-3-16 Jonan, Yonezawa 992-8510 (Japan)

2016-11-30

Highlights: • RT-ALD of ZrO{sub 2} is developed using TEMAZ and plasma-excited humidified argon. • The plasma-excited humidified argon is effective in oxidizing the TEMAZ saturated ZrO{sub 2}. • We discuss the reaction mechanism of the RT-ZrO{sub 2} ALD. - Abstract: Room-temperature atomic layer deposition (ALD) of ZrO{sub 2} is developed with tetrakis(ethylmethylamino)zirconium (TEMAZ) and a plasma-excited humidified argon. A growth per cycle of 0.17 nm/cycle at room temperature is confirmed, and the TEMAZ adsorption and its oxidization on ZrO{sub 2} are characterized by IR absorption spectroscopy with a multiple internal reflection mode. TEMAZ is saturated on a ZrO{sub 2} surface with exposures exceeding ∼2.0 × 10{sup 5} Langmuir (1 Langmuir = 1.0 × 10{sup −6} Torr s) at room temperature, and the plasma-excited humidified argon is effective in oxidizing the TEMAZ-adsorbed ZrO{sub 2} surface. The IR absorption spectroscopy suggests that Zr-OH works as an adsorption site for TEMAZ. The reaction mechanism of room-temperature ZrO{sub 2} ALD is discussed in this paper.

Wear and chemistry of zirconium-silicate, aluminium-silicate and zirconium-aluminium-silicate glasses in alkaline medium

International Nuclear Information System (INIS)

Rouse, C.G.; Lemos Guenaga, C.M. de

1984-01-01

A study of the chemical durability, in alkaline solutions, of zirconium silicate, aluminium silicate, zirconium/aluminium silicate glasses as a function of glass composition is carried out. The glasses were tested using standard DIN-52322 method, where the glass samples are prepared in small polished pieces and attacked for 3 hours in a 800 ml solution of 1N (NaOH + NA 2 CO 3 ) at 97 0 C. The results show that the presence of ZrO 2 in the glass composition increases its chemical durability to alkaline attack. Glasses of the aluminium/zirconium silicate series were melted with and without TiO 2 . It was shown experimentally that for this series of glasses, the presence of both TiO 2 and ZrO 2 gave better chemical durability results. However, the best overall results were obtained from the simpler zirconium silicate glasses, where it was possible to make glasses with higher values of ZrO 2 . (Author) [pt

Fuel assembly for FBR type reactor and reactor core thereof

International Nuclear Information System (INIS)

Kobayashi, Kaoru.

1998-01-01

The present invention provides a fuel assembly to be loaded to a reactor core of a large sized FBR type reactor, in which a coolant density coefficient can be reduced without causing power peaking in the peripheral region of neutron moderators loaded in the reactor core. Namely, the fuel assembly for the FBR type reactor comprises a plurality of fission product-loaded fuel rods and a plurality of fertile material-loaded fuel rods and one or more rods loading neutron moderators. In this case, the plurality of fertile material-loaded fuel rods are disposed to the peripheral region of the neutron moderator-loaded rods. The plurality of fission product-loaded fuel rods are disposed surrounding the peripheral region of the plurality of fertile material-loaded fuel rods. The neutron moderator comprises zirconium hydride, yttrium hydride and calcium hydride. The fission products are mixed oxide fuels. The fertile material comprises depleted uranium or natural uranium. (I.S.)

Study for the chlorination of zirconium oxide

International Nuclear Information System (INIS)

Seo, E.S.M.; Takiishi, H.; Paschoal, J.O.A.; Andreoli, M.

1990-12-01

In the development of new ceramic and metallic materials the chlorination process constitutes step in the formation of several intermediate compounds, such as metallic chlorides, used for the production of high, purity raw materials. Chlorination studies with the aim of fabrication special zirconium-base alloys have been carried out at IPEN. Within this program the chlorination technique has been used for zirconium tetrachloride production from zirconium oxide. In this paper some relevant parameters such as: time and temperature of reaction, flow rate of chloride gas and percentage of the reducing agent which influence the efficiency of chlorination of zirconium oxide are evaluated. Thermodynamical aspects about the reactions involved in the process are also presented. (author)

Ductile zirconium powder by hydride-dehydride process

Energy Technology Data Exchange (ETDEWEB)

Krishnan, T S [BHABHA ATOMIC RESEARCH CENTRE, BOMBAY (INDIA); CHAUDHARY, S [NUCLEAR FUEL COMPLEX, HYDERABAD (INDIA)

1976-09-01

The preparation of ductile zirconium powder by the hydride-dehydride process has been described. In this process massive zirconium obtained from Kroll reduction of ZrCl/sub 4/ is first rendered brittle by hydrogenation and the hydride crushed and ground in a ball mill to the required particle size. Hydrogen is then hot vacuum extracted to yield the metal powder. The process has been successfully employed for the production of zirconium powders with low oxygen content and having hardness values in the range of 115-130 BHN, starting from a zirconium sponge of 100-120 BHN hardness. Influence of surface characteristics of the starting metal on its hydriding behaviour has been studied and the optimum hydriding-dehydriding conditions established.

Joint titrimetric determination of zirconium and hafnium

International Nuclear Information System (INIS)

Vazquez, Cristina; Botbol, Moises; Bianco de Salas, G.N.; Cornell de Casas, M.I.

1980-01-01

A method for the joint titrimetric determination of zirconium and hafnium, which are elements of similar chemical behaviour, is described. The disodic salt of the ethylendiaminetetracetic acid (EDTA) is used for titration, while xilenol orange serves as final point indicator. Prior to titration it is important to evaporate with sulfuric acid, the solution resulting from the zirconium depolymerization process, to adjust the acidity and to eliminate any interferences. The method, that allows the quick and precise determination of zirconium and hafnium in quantities comprised between 0.01 and mg, was applied to the analysis of raw materials and of intermediate and final products in the fabrication of zirconium sponge and zircaloy. (M.E.L.) [es

Investigation of Zirconium Oxide Films in Different Dissolved Hydrogen Concentration

International Nuclear Information System (INIS)

Kim, Taeho; Choi, Kyoung Joon; Yoo, Seung Chang; Kim, Ji Hyun

2016-01-01

It has been reported that in pre-transition zirconium oxide, the volume fraction of tetragonal zirconium oxide increased near the oxide/metal (O/M) interface, and the sub-stoichiometric zirconium oxide layer was observed. The diffusion of oxygen ion through the oxide layer is the rate-limiting process during the pre-transition oxidation process, and this diffusion mainly occurs in the grain boundaries. The two layered oxide structure is formed in pre-transition oxide for the zirconium alloy in high-temperature water environment. It is known that the corrosion rate is related to the volume fraction of zirconium oxide and the pores in the oxides; therefore, the aim of this paper is to investigate the oxidation behavior in the pre-transition zirconium oxide in high-temperature water chemistry. In this study, in situ Raman and TEM analysis were conducted for investigating the phase transformation of zirconium alloy in primary water. From this study, the following conclusions are drawn: 1. The zirconium alloy was oxidized in primary water chemistry for 100 d, and Raman and TEM were measured after 30, 50, 80, and 100 d from start-up. 2. TEM and FFT analysis showed that the zirconium oxide mostly consisted of the monoclinic phase. The tetragonal zirconium oxide was just found near the O/M interface

Method of operating a water-cooled nuclear reactor

International Nuclear Information System (INIS)

Lysell, G.

1975-01-01

When operating a water-cooled nuclear reactor, in which the fuel rods consist of zirconium alloy tubes containing an oxidic nuclear fuel, stress corrosion in the tubes can be reduced or avoided if the power of the reactor is temporarily increased so much that the thermal expansion of the nuclear fuel produces a flow of the material in the tube. After that temporary power increase the power output is reduced to the normal power

Ternary cobalt-molybdenum-zirconium coatings for alternative energies

Science.gov (United States)

Yar-Mukhamedova, Gulmira; Ved', Maryna; Sakhnenko, Nikolay; Koziar, Maryna

2017-11-01

Consistent patterns for electrodeposition of Co-Mo-Zr coatings from polyligand citrate-pyrophosphate bath were investigated. The effect of both current density amplitude and pulse on/off time on the quality, composition and surface morphology of the galvanic alloys were determined. It was established the coating Co-Mo-Zr enrichment by molybdenum with current density increasing up to 8 A dm-2 as well as the rising of pulse time and pause duration promotes the content of molybdenum because of subsequent chemical reduction of its intermediate oxides by hydrogen ad-atoms. It was found that the content of the alloying metals in the coating Co-Mo-Zr depends on the current density and on/off times extremely and maximum Mo and Zr content corresponds to the current density interval 4-6 A dm-2, on-/off-time 2-10 ms. Chemical resistance of binary and ternary coatings based on cobalt is caused by the increased tendency to passivity and high resistance to pitting corrosion in the presence of molybdenum and zirconium, as well as the acid nature of their oxides. Binary coating with molybdenum content not less than 20 at.% and ternary ones with zirconium content in terms of corrosion deep index are in a group ;very proof;. It was shown that Co-Mo-Zr alloys exhibits the greatest level of catalytic properties as cathode material for hydrogen electrolytic production from acidic media which is not inferior a platinum electrode. The deposits Co-Mo-Zr with zirconium content 2-4 at.% demonstrate high catalytic properties in the carbon(II) oxide conversion. This confirms the efficiency of materials as catalysts for the gaseous wastes purification and gives the reason to recommend them as catalysts for red-ox processes activating by oxygen as well as electrode materials for red-ox batteries.

Voltammetric determination of zirconium using azo compounds

International Nuclear Information System (INIS)

Orshulyak, O.O.; Levitskaya, G.D.

2008-01-01

The optimum conditions for zirconium complexation with azo compounds are found. The applicability of Eriochrome Red B, Calcon, and Calcion to the voltammetric determination of zirconium, total Zr(IV) and Hf(IV), and Zr(IV) in the presence of Zn(II), Cu(II), Cd(II), Ni(II), or Ti(IV) is demonstrated. The developed procedures are used to determine zirconium in a terbium alloy and in an alloy for airplane wheel drums [ru

Modelling of zirconium alloys corrosion in LWRs

International Nuclear Information System (INIS)

Kritskij, V.G.; Berezina, I.G.; Kritskij, A.V.; Stjagkin, P.S.

1999-01-01

Chemical parameters, that exerted effect on Zr+1%Nb alloy corrosion and deserved consideration during reactor operation, were defined and a model was developed to describe the influence of physical and chemical parameters on zirconium alloys corrosion in nuclear power plants. The model is based on the correlation between the zirconium oxide solubility in high-temperature water under the influence of the chemical parameters and the measured values of fuel cladding corrosion under LWR conditions. The intensity of fuel cladding corrosion in the primary circuits depends on the coolant water quality, growth of iron oxide deposits and vaporization portion. Mathematically, the oxidation rate can be expressed as a sum of heat and radiation components. The temperature dependence on the oxidation rate can be described by the Arrenius equation. The radiation component of Zr uniform corrosion equation is a function of several factors such as neutron fluency, the temperature the metallurgical composition and et. We assume that the main factor is the changing of water chemistry and the H 2 O 2 concentration play the determinative role. Probably, the influence of H 2 O 2 is based on the formation of unstable compound ZrO 3 ·nH 2 O and Zr(OH) 4 with high solubility. The validity of the used formulae was confirmed by corrosion measurements on WWER and RBMK fuel cladding. The model can be applied for calculating the reliability of nuclear fuel operation. (author)

A contribution to the study of the mixed uranium-plutonium mono-carbides containing small quantities of zirconium; Contribution a l'etude du monocarbure d'uranium et de plutonium avec de faibles additions de zirconium

Energy Technology Data Exchange (ETDEWEB)

Bocker, S [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1970-03-01

We have studied a mixed monocarbide, type (U,Pu)C, containing small additions of zirconium for the application as a fast neutron reactor fuel. A preliminary study was conducted on the (U,Zr)C monocarbide (Report CEA-R-3765(1). It was found that small additions of zirconium to the uranium-plutonium monocarbide improve a number of properties such as atmospheric corrosion, the hardness, and particularly the compatibility with 316 stainless steel. However, properties such as the coefficient of expansion and the melting point are only slightly changed. The relative percentage of Pu/U+Pu in the monocarbide was fixed at 20 per cent. Two processes of fabrication were employed: casting in an arc furnace, sintering, carried out after having the hydrides of the metals carburized. The metallurgical results indicate, that the above mentioned fuel might be of interest for fast neutron reactor application. (author) [French] On a etudie un combustible de type carbure (U,Pu)C pour les reacteurs a neutrons rapides. Les recherches preliminaires ont porte sur le carbure (UZr)C (rapport CEA-R-3765(1)). L'addition de faibles quantites de zirconium (3 at. pour cent) au monocarbure (U,Pu)C, ameliore certaines proprietes, commee la tenue a la corrosion atmospherique, la durete et surtout la compatibilite avec l'acier inoxydable X-18 M, Par contre le coefficient de dilatation et la densite sont peu changes. Le rapport Pu/Pu+U etait fixe a 20 pour cent. Deux procedes de fabrication ont ete etudies: l'un par fusion a l'arc, l'autre par frittage a partir de metaux hydrures. Au vu des resultats metallurgiques obtenus le carbure (U,Pu,Zr)C semble presenter un interet certain. (auteur)

Zirconium cladding - the long way towards a mechanistic understanding of processing and performance

International Nuclear Information System (INIS)

Preuss, Michael

2011-01-01

Zirconium alloys are the material of choice to encapsulate nuclear fuel in light and heavy water-cooled reactors due to their low neutron absorption, excellent corrosion resistance and sufficient mechanical properties. Despite these advantageous physical and mechanical properties a more physically based understanding of microstructure and texture evolution during processing is highly desirable in order to improve our understanding of formability during thermomechanical processing and performance variability of cladding material. In addition, the purely empirical understanding of aqueous zirconium corrosion, hydrogen pick up, hydride precipitation as well as irradiation growth and creep limits the accuracy of life predictions and therefore the level of burnup that is obtained from current fuel assemblies. The presentation aims at giving examples of new research strategies that will enable the development of a new physical understanding of processing and performance aspects in zirconium cladding material, which is required to develop new predictive models. Particular emphasis will be placed on using novel research tools and large-scale research facilities such as neutron spallation and synchrotron radiation sources to undertake very detailed and often in-situ studies of deformation mechanisms and microstructure evolution as well as determining stress states in grain families, oxides and hydrides. The results will be presented in the view of how they might help us to improve our understanding and enable the development of better predictive models

URANIUM DECONTAMINATION WITH RESPECT TO ZIRCONIUM

Science.gov (United States)

Vogler, S.; Beederman, M.

1961-05-01

A process is given for separating uranium values from a nitric acid aqueous solution containing uranyl values, zirconium values and tetravalent plutonium values. The process comprises contacting said solution with a substantially water-immiscible liquid organic solvent containing alkyl phosphate, separating an organic extract phase containing the uranium, zirconium, and tetravalent plutonium values from an aqueous raffinate, contacting said organic extract phase with an aqueous solution 2M to 7M in nitric acid and also containing an oxalate ion-containing substance, and separating a uranium- containing organic raffinate from aqueous zirconium- and plutonium-containing extract phase.

NSRR experiment with un-irradiated uranium-zirconium hydride fuel. Design, fabrication process and inspection data of test fuel rod

Energy Technology Data Exchange (ETDEWEB)

Sasajima, Hideo; Fuketa, Toyoshi; Ishijima, Kiyomi; Kuroha, Hiroshi; Ikeda, Yoshikazu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Aizawa, Keiichi

1998-08-01

An experiment plan is progressing in the Nuclear Safety Research Reactor (NSRR) to perform pulse-irradiation with uranium-zirconium hydride (U-ZrH{sub x}) fuel. This fuel is widely used in the training research and isotope production reactor of GA (TRIGA). The objectives of the experiment are to determine the fuel rod failure threshold and to investigate fuel behavior under simulated reactivity initiated accident (RIA) conditions. This report summarizes design, fabrication process and inspection data of the test fuel rods before pulse-irradiation. The experiment with U-ZrH{sub x} fuel will realize precise safety evaluation, and improve the TRIGA reactor performance. The data to be obtained in this program will also contribute development of next-generation TRIGA reactor and its safety evaluation. (author)

Structural and luminescence properties of europium(III)-doped zirconium carbonates and silica-supported Eu3+-doped zirconium carbonate nanoparticles

International Nuclear Information System (INIS)

Sivestrini, S.; Riello, P.; Freris, I.; Cristofori, D.; Enrichi, F.; Benedetti, A.

2010-01-01

The synthesis, morphology and luminescence properties of europium(III)-doped zirconium carbonates prepared as bulk materials and as silica-supported nanoparticles with differing calcination treatments are reported. Transmission electron microscopy and X-ray diffraction analyses have, respectively, been used to study the morphology and to quantify the atomic amount of europium present in the optically active phases of the variously prepared nanomaterials. Rietveld analysis was used to quantify the constituting phases and to determinate the europium content. Silica particles with an approximate size of 30 nm were coated with 2 nm carbonate nanoparticles, prepared in situ on the surface of the silica core. Luminescence measurements revealed the role of different preparation methods and of europium-doping quantities on the optical properties observed.

Atomic layer deposition on nanoparticles in a rotary reactor

Science.gov (United States)

McCormick, Jarod Alan

Challenges are encountered during atomic layer deposition (ALD) on large quantities of nanoparticles. The particles must be agitated or vigorously mixed to perform the ALD surface reactions in reasonable times and to prevent the particles from being agglomerated by the ALD film. The high surface area of nanoparticles also demands efficient reactant usage because large quantities of reactant are required for the surface reactions to reach completion. To address these challenges, a novel rotary reactor was developed to achieve constant particle agitation during static ALD reactant exposures. In the design of this new reactor, a cylindrical drum with porous metal walls was positioned inside a vacuum chamber. The porous cylindrical drum was rotated by a magnetically coupled rotary feedthrough. By rotating the cylindrical drum to obtain a centrifugal force of less than one gravitational force, the particles were agitated by a continuous "avalanche" of particles. The effectiveness of this rotary reactor was demonstrated by Al 2O3 ALD on ZrO2 particles. A number of techniques including transmission electron microscopy, Fourier transform infrared spectroscopy, scanning Auger spectroscopy and x-ray photoelectron spectroscopy confirmed that the Al2O3 ALD film conformally coats the ZrO 2 particles. Combining static reactant exposures with a very high surface area sample in the rotary reactor also provides unique opportunities for studying the surface chemistry during ALD. Sequential, subsaturating doses can be used to examine the self-limiting behavior of the ALD reactions in the rotary reactor. This dosing method is the first demonstration of self-limiting ALD on bulk quantities of nanoparticles. By combining these sequential, subsaturating doses with quadrupole mass spectrometry, ALD reactions can be analyzed from the gas phase using full mass spectrum analysis. The reaction products are present in a high enough concentration to discern a gas phase mechanism for reactions

The Atomic energy basic law

International Nuclear Information System (INIS)

1979-01-01

The law aims to secure future energy resources, push forward progress of science and advancement of industry for welfare of the mankind and higher standard of national life by helping research, development and utilization of atomic power. Research, development and utilization of atomic power shall be limited to the peaceful purpose with emphasis laid on safety and carried on independently under democratic administration. Basic concepts and terms are defined, such as: atomic power; nuclear fuel material; nuclear raw material; reactor and radiation. The Atomic Energy Commission and the Atomic Energy Safety Commission shall be set up at the Prime Minister's Office deliberately to realize national policy of research, development and utilization of atomic power and manage democratic administration for atomic energy. The Atomic Energy Commission shall plan, consider and decide matters concerning research, development and utilization of atomic energy. The Atomic Energy Safety Commission shall plan, consider and decide issues particularly concerning safety securing among such matters. The Atomic Energy Research Institute shall be founded under the governmental supervision to perform research, experiment and other necessary affairs for development of atomic energy. The Power Reactor and Nuclear Fuel Development Corporation shall be established likewise to develop fast breeding reactor, advanced thermal reactor and nuclear fuel materials. Development of radioactive minerals, control of nuclear fuel materials and reactors and measures for patent and invention concerning atomic energy, etc. are stipulated respectively. (Okada, K.)

Nuclear science and metallurgy. Advances and interactions

Energy Technology Data Exchange (ETDEWEB)

Grison, E [CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

1977-01-01

The history of the production of atomic power by fission since 1953 is reviewed: metallurgy of uranium; the new metals zirconium and beryllium; steels and nuclear structure; nuclear reactor vessels; water corrosion; effects of radiations.

Modification of Different Zirconium Propoxide Precursors by Diethanolamine. Is There a Shelf Stability Issue for Sol-Gel Applications?

Science.gov (United States)

Spijksma, Gerald I.; Blank, Dave H. A.; Bouwmeester, Henny J. M.; Kessler, Vadim G.

2009-01-01

Modification of different zirconium propoxide precursors with H2dea was investigated by characterization of the isolated modified species. Upon modification of zirconium n-propoxide and [Zr(OnPr)(OiPr)3(iPrOH)]2 with ½ a mol equivalent of H2dea the complexes [Zr2(OnPr)6(OCH2CH2)2NH]2 (1) and [Zr2(OnPr)2(OiPr)4(OCH2CH2)2NH]2 (2) were obtained. However, 1H-NMR studies of these tetranuclear compounds showed that these are not time-stable either in solution or solid form. The effect of this time instability on material properties is demonstrated by light scattering and TEM experiments. Modification of zirconium isopropoxide with either ½ or 1 equivalent mol of H2dea results in formation of the trinuclear complex, Zr{η3μ2-NH(C2H4O)2}3[Zr(OiPr)3]2(iPrOH)2 (3) countering a unique nona-coordinated central zirconium atom. This complex 3 is one of the first modified zirconium propoxide precursors shown to be stable in solution for long periods of time. The particle size and morphology of the products of sol-gel synthesis are strongly dependent on the time factor and eventual heat treatment of the precursor solution. Reproducible sol-gel synthesis requires the use of solution stable precursors. PMID:20087472

Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Smith, Kurt R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-09-01

The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designs allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.

Growth and characterization of oxide layers on zirconium alloys

International Nuclear Information System (INIS)

Maroto, A.J.G.; Bordoni, R.; Villegas, M.; Blesa, M.A.; Olmedo, A.M.; Iglesias, A.; Rigotti, G.

1997-01-01

Corrosion behaviour in aqueous media at high temperature of zirconium alloys has been extensively studied in order to elucidate the corrosion mechanism and kinetics. The characterization of the morphology and microstructure of these oxides through the different stages of oxide growth may contribute to understand their corrosion mechanism. Argentina has initiated a research program to correlate long term in and out-reactor corrosion of these alloys. This paper reports a comparative study of out of pile oxidation of Zr-2.5Nb and Zry-4, which are structural materials of in-core components of nuclear power plants. Kinetic data at different temperatures and microstructural characterization of the oxide films are presented. (author). 25 refs, 18 figs, 1 tab

Growth and characterization of oxide layers on zirconium alloys

Energy Technology Data Exchange (ETDEWEB)

Maroto, A J.G.; Bordoni, R; Villegas, M; Blesa, M A; Olmedo, A M; Iglesias, A; Rigotti, G [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

1997-02-01

Corrosion behaviour in aqueous media at high temperature of zirconium alloys has been extensively studied in order to elucidate the corrosion mechanism and kinetics. The characterization of the morphology and microstructure of these oxides through the different stages of oxide growth may contribute to understand their corrosion mechanism. Argentina has initiated a research program to correlate long term in and out-reactor corrosion of these alloys. This paper reports a comparative study of out of pile oxidation of Zr-2.5Nb and Zry-4, which are structural materials of in-core components of nuclear power plants. Kinetic data at different temperatures and microstructural characterization of the oxide films are presented. (author). 25 refs, 18 figs, 1 tab.

The ReactorSTM: Atomically resolved scanning tunneling microscopy under high-pressure, high-temperature catalytic reaction conditions

Energy Technology Data Exchange (ETDEWEB)

Herbschleb, C. T.; Tuijn, P. C. van der; Roobol, S. B.; Navarro, V.; Bakker, J. W.; Liu, Q.; Stoltz, D.; Cañas-Ventura, M. E.; Verdoes, G.; Spronsen, M. A. van; Bergman, M.; Crama, L.; Taminiau, I.; Frenken, J. W. M., E-mail: frenken@physics.leidenuniv.nl [Huygens-Kamerlingh Onnes Laboratory, Leiden University, P.O. box 9504, 2300 RA Leiden (Netherlands); Ofitserov, A.; Baarle, G. J. C. van [Leiden Probe Microscopy B.V., J.H. Oortweg 21, 2333 CH Leiden (Netherlands)

2014-08-15

To enable atomic-scale observations of model catalysts under conditions approaching those used by the chemical industry, we have developed a second generation, high-pressure, high-temperature scanning tunneling microscope (STM): the ReactorSTM. It consists of a compact STM scanner, of which the tip extends into a 0.5 ml reactor flow-cell, that is housed in a ultra-high vacuum (UHV) system. The STM can be operated from UHV to 6 bars and from room temperature up to 600 K. A gas mixing and analysis system optimized for fast response times allows us to directly correlate the surface structure observed by STM with reactivity measurements from a mass spectrometer. The in situ STM experiments can be combined with ex situ UHV sample preparation and analysis techniques, including ion bombardment, thin film deposition, low-energy electron diffraction and x-ray photoelectron spectroscopy. The performance of the instrument is demonstrated by atomically resolved images of Au(111) and atom-row resolution on Pt(110), both under high-pressure and high-temperature conditions.

Translucency and Strength of High-Translucency Monolithic Zirconium-Oxide Materials

Science.gov (United States)

2016-05-12

Capt Todd D. Church APPROVED: Translucency and Strength of High-Translucency Monolithic Zirconium -Oxide Materials C~t) Kraig/[ Vandewalle Date...copyrighted material in the thesis/dissertation manuscript entitled: "Translucency arid Strength of High-Translucency Monolithic Zirconium -Oxide...Translucency Monolithic Zirconium -Oxide Materials Abstract Dental materials manufacturers have developed more translucent monolithic zirconium oxide

Review of zirconium-zircaloy pyrophoricity

International Nuclear Information System (INIS)

Cooper, T.D.

1984-11-01

Massive zirconium metal scrap can be handled, shipped, and stored with no evidence of combustion or pyrophoricity hazards. Mechanically produced fine scrap such as shavings, turnings, or powders can burn but are not pyrophoric unless the particle diameter is less than 54 μm. Powders with particle diameters less than 54 μm can be both pyrophoric and explosive. Pyrophoric powders should be collected and stored underwater or under inert gas cover to reduce the flammability hazard. Opening sealed containers of zirconium stored underwater should be attempted with caution since hydrogen may be present. The factors that influence the ignition temperature have been explored in depth and recommendations are included for the safe handling, shipping, and storage of pyrophoric or flammable zirconium. 29 refs., 5 figs., 6 tabs

Tests for depositing thin films of metallic zirconium; Essais de depot de zirconium metallique en couches minces

Energy Technology Data Exchange (ETDEWEB)

Bentolila, J.; Pattoret, A.; Platzer, R.

1957-01-15

The authors report a study which aimed at obtaining a thin, adhesive and non porous coating of metallic zirconium on a uranium substrate by means of chemical process. The main required condition was not to go beyond the uranium phase change temperature (650 C). Two kinds of tests have been performed: on the one hand, tests of reduction of zirconium tetrachloride in non aqueous solvent medium, and on the other hand, tests of vacuum decomposition of zirconium hydride. As far as the first tests are concerned, the authors studied organic solvent media (reduction by aluminium and lithium hydride, action of organic-magnesium compounds), and liquid ammoniac. For the second test type, they describe the apparatus, the preparation of the zirconium hydride, preparation of the substrate surfaces, coating preparation, and decomposition process. Results are discussed in terms of temperature, of presence of copper powder in the coating, of early surface hydriding of uranium, surface polishing.

Removal of iron contaminant from zirconium chloride solution

International Nuclear Information System (INIS)

Voit, D.O.

1992-01-01

This patent describes a process for eliminating iron contaminant from an aqueous zirconium chloride solution that has been contaminated with FeCl 3 in a plant in which zirconium and hafnium chloride solutions are separated by a main MINK solvent extraction system and the FeCl 3 is normally removed from the zirconium chloride solution by a secondary MINK solvent extraction system

Nuclear science and metallurgy. Advances and interactions

International Nuclear Information System (INIS)

Grison, Emmanuel

1977-01-01

The history of the production of atomic power by fission since 1953 is reviewed: metallurgy of uranium; the new metals zirconium and beryllium; steels and nuclear structure; nuclear reactor vessels; water corrosion; effects of radiations [fr

Maintenance experience on reactor recirculation pumps at Tarapur Atomic Power Station

International Nuclear Information System (INIS)

Singh, A.K.

1995-01-01

Reactor recirculation pumps at Tarapur Atomic Power Station (TAPS) are vertical, single stage centrifugal pumps having mechanical shaft seals and are driven by vertical mounted 3.3 kV, 3 phase, 1500 h.p. electric motors. During these years of operation TAPS has gained enough experience and expertise on the maintenance of reactor recirculation pumps which are dealt in this article. Failure of mechanical shaft seals, damage on pump carbon bearings, motor winding insulation failures and motor shaft damage have been the main areas of concern on recirculation pump. A detailed procedure step by step with component sketches has helped in eliminating errors during shaft seal assembly and installation. Pressure breakdown devices in seal assembly were rebuilt. Additional coolant water injection for shaft seal cooling was provided. These measures have helped in extending the reactor recirculation pump seal life. Pump bearing problems were mainly due to failure of anti-rotation pins and dowel pins of bearing assembly. These pins were redesigned and strengthened. Motor stator winding insulation failures were detected. Stator winding replacement program has been taken up on regular basis to avoid winding insulation failure due to aging. 3 refs., 2 tabs., 7 figs

High purity zirconium obtainment through the iodine compounds transport method

International Nuclear Information System (INIS)

Bolcich, J.C.; Zuzek, E.; Dutrus, S.M.; Corso, H.L.

1987-01-01

This paper describes the experimental method and the equipment designed, constructed and actually applied for the high purity zirconium obtainment from a zirconium sponge of the nuclear type. The mechanism of purification is based on the impure metal attack with gaseous iodine (at 200 deg C) to obtain zirconium tetra iodine as main product which is then transformed into a pure zirconium base (at 1000-1300 deg C), precipitating the metallic zirconium and releasing the gaseous iodine. From the first experiences carried out, pure zirconium has been obtained from an initial filament of 0.5 mm of diameter as well as wires up to 2.5 mm of diameter. This work presents the results from the studies and analysis made to characterize the material obtained. Finally, the refining methods to which the zirconium produced may be submitted so as to optimize the final purity are discussed. (Author)

Histomorphometric and histologic evaluation of titanium-zirconium (aTiZr) implants with anodized surfaces.

Science.gov (United States)

Sharma, Ajay; McQuillan, A James; Shibata, Yo; Sharma, Lavanya A; Waddell, John Neil; Duncan, Warwick John

2016-05-01

The choice of implant surface has a significant influence on osseointegration. Modification of TiZr surface by anodization is reported to have the potential to modulate the osteoblast cell behaviour favouring more rapid bone formation. The aim of this study is to investigate the effect of anodizing the surface of TiZr discs with respect to osseointegration after four weeks implantation in sheep femurs. Titanium (Ti) and TiZr discs were anodized in an electrolyte containing DL-α-glycerophosphate and calcium acetate at 300 V. The surface characteristics were analyzed by scanning electron microscopy, electron dispersive spectroscopy, atomic force microscopy and goniometry. Forty implant discs with thickness of 1.5 and 10 mm diameter (10 of each-titanium, titanium-zirconium, anodized titanium and anodized titanium-zirconium) were placed in the femoral condyles of 10 sheep. Histomorphometric and histologic analysis were performed 4 weeks after implantation. The anodized implants displayed hydrophilic, porous, nano-to-micrometer scale roughened surfaces. Energy dispersive spectroscopy analysis revealed calcium and phosphorous incorporation into the surface of both titanium and titanium-zirconium after anodization. Histologically there was new bone apposition on all implanted discs, slightly more pronounced on anodised discs. The percentage bone-to-implant contact measurements of anodized implants were higher than machined/unmodified implants but there was no significant difference between the two groups with anodized surfaces (P > 0.05, n = 10). The present histomorphometric and histological findings confirm that surface modification of titanium-zirconium by anodization is similar to anodised titanium enhances early osseointegration compared to machined implant surfaces.

A recent investigation of neutron total cross section of zirconium in the wavelength range (0.1-1.25) Ao

International Nuclear Information System (INIS)

Abu El-Ela, M.A.

1996-01-01

The neutron total cross section of zirconium has been investigated in the neutron wavelength range (0.1 -1.52) A o by using slow neutron time of flight spectrometer, installed in front of the horizontal channel No.6 of the ETRR-1 reactor (2MW). The results have showed that the neutrons with short wavelength (0.1 - 0.76) A o cannot interact with the crystal structure while it can interact with the free bound atom to give the value (6.2 +0.1) barns for the potential scattering cross section or (the scattering length = 6.2 fermi)). The present measured value is in good agreement with the international published values by different technique. The neutrons with longer wavelength (0.76 - 1.52) A o have showed dependence of the total cross section on the neutron wavelength. Such dependence between the total cross section and the neutron wavelength can not be observed in the reported previous measurements, which can be attributed to the limited number of the measured values. 4 figs

Quercetin as colorimetric reagent for determination of zirconium

Science.gov (United States)

Grimaldi, F.S.; White, C.E.

1953-01-01

Methods described in the literature for the determination of zirconium are generally designed for relatively large amounts of this element. A good procedure using colorimetric reagent for the determination of trace amounts is desirable. Quercetin has been found to yield a sensitive color reaction with zirconium suitable for the determination of from 0.1 to 50?? of zirconium dioxide. The procedure developed involves the separation of zirconium from interfering elements by precipitation with p-dimethylaminoazophenylarsonic acid prior to its estimation with quercetin. The quercetin reaction is carried out in 0.5N hydrochloric acid solution. Under the operating conditions it is indicated that quercetin forms a 2 to 1 complex with zirconium; however, a 2 to 1 and a 1 to 1 complex can coexist under special conditions. Approximate values for the equilibrium constants of the complexes are K1 = 0.33 ?? 10-5 and K2 = 1.3 ?? 10-9. Seven Bureau of Standards samples of glass sands and refractories were analyzed with excellent results. The method described should find considerable application in the analysis of minerals and other materials for macro as well as micro amounts of zirconium.

Antimony removal from aqueous solutions using Zirconium hydroxide

International Nuclear Information System (INIS)

Petrescu, D.; Velciu, L.; Bucur, C.

2016-01-01

In this paper it is presented an experimental test for non-radioactive antimony removal from aqueous solutions using zirconium hydroxide powder. Also, it was studied how the temperature and pH influences antimony adsorption onto zirconium hydroxide surface. After the adsorption, solutions were filtered on Cellulose Mixed Ester Membrane with 0.2 μm pore size to remove the zirconium powder and then the aqueous solutions were sent to Inductively Coupled Plasma Optic Emission Spectrometry (ICP-OES) for quantitative analysis of Sb. Zirconium hydroxide powders were examined by optical microscopy. For the solutions that were tested at pH 4.5 and 10.2 the antimony concentration dropped below the detection limit of ICP-OES device, proof of antimony adsorption on zirconium hydroxide. Also, for the other tested solutions which had pH=12 the antimony concentration reduced with 77% and 80%. The temperature had no influence upon adsorption mechanism. (authors)

Structural, morphological and dielectric studies of zirconium substituted CoFe2O4 nanoparticles

Directory of Open Access Journals (Sweden)

S. Anand

2017-12-01

Full Text Available In this work, the influence of zirconium substitution in cubic spinel nanocrystalline CoFe2O4 on the structural, morphological and dielectric properties are reported. Zirconium substituted cobalt ferrite Co1-xZrxFe2O4 (x = 0.7 nanoparticles were synthesized by sol-gel route. The structural and morphological investigations using powder X-ray diffraction and high resolution scanning electron microscope (HRSEM analysis are reported. Scherrer plot, Williamson–Hall analysis and Size-strain plot method were used to calculate the crystallite size and lattice strain of the samples. High purity chemical composition of the sample was confirmed by energy dispersive X-ray analysis. The atoms vibration modes of as synthesized nanoparticles were recorded using Fourier transform infrared (FTIR spectrometer in the range of 4000–400 cm-1. The temperature-dependent dielectric properties of zirconium substituted cobalt ferrite nanoparticles were also carried out. Relative dielectric permittivity, loss tangent and AC conductivity were measured in the frequency range 50 Hz to 5 MHz at temperatures between 323 K and 473 K. The dielectric constant and dielectric loss values of the sample decreased with increasing in the frequency of the applied signal.

Molten salt extractive distillation process for zirconium-hafnium separation

International Nuclear Information System (INIS)

McLaughlin, D.F.; Stoltz, R.A.

1989-01-01

This patent describes an improvement in a process for zirconium-hafnium separation. It utilizes an extractive distillation column with a mixture of zirconium and hafnium tetrachlorides introduced into a distillation column having a top and bottom with hafnium enriched overheads taken from the top of the column and a molten salt solvent circulated through the column to provide a liquid phase, and with molten salt solvent containing zirconium chloride being taken from the bottom of the distillation column. The improvements comprising: utilizing a molten salt solvent consisting principally of lithium chloride and at least one of sodium, potassium, magnesium and calcium chlorides; stripping of the zirconium chloride taken from the bottom of the distillation column by electrochemically reducing zirconium from the molten salt solvent; and utilizing a pressurized reflux condenser on the top of the column to add the hafnium chloride enriched overheads to the molten salt solvent previously stripped of zirconium chloride

Proceedings of the nineteenth symposium of atomic energy research on WWER reactor physics and reactor safety

International Nuclear Information System (INIS)

Vidovszky, I.

2009-10-01

The present volume contains 55 papers, presented on the nineteenth symposium of atomic energy research, held in Varna, Bulgaria, 21-25 September 2009. The papers are presented in their original form, i. e. no corrections or modifications were carried out. The content of this volume is divided into thematic groups: Fuel Management, Spectral and Core Calculations, Core Surveillance and Monitoring, CFD Analysis, Reactor Dynamics Thermal Hydraulics and Safety Analysis, Physical Problems of Spent Fuel Decommissioning and Radwaste, Actinide Transmutation and Spent Fuel Disposal, Core Operation, Experiments and Code Validation - according to the presentation sequence on the Symposium. (Author)

A contribution to the study of the mixed uranium-plutonium mono-carbides containing small quantities of zirconium; Contribution a l'etude du monocarbure d'uranium et de plutonium avec de faibles additions de zirconium

Energy Technology Data Exchange (ETDEWEB)

Bocker, S. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1970-03-01

We have studied a mixed monocarbide, type (U,Pu)C, containing small additions of zirconium for the application as a fast neutron reactor fuel. A preliminary study was conducted on the (U,Zr)C monocarbide (Report CEA-R-3765(1). It was found that small additions of zirconium to the uranium-plutonium monocarbide improve a number of properties such as atmospheric corrosion, the hardness, and particularly the compatibility with 316 stainless steel. However, properties such as the coefficient of expansion and the melting point are only slightly changed. The relative percentage of Pu/U+Pu in the monocarbide was fixed at 20 per cent. Two processes of fabrication were employed: casting in an arc furnace, sintering, carried out after having the hydrides of the metals carburized. The metallurgical results indicate, that the above mentioned fuel might be of interest for fast neutron reactor application. (author) [French] On a etudie un combustible de type carbure (U,Pu)C pour les reacteurs a neutrons rapides. Les recherches preliminaires ont porte sur le carbure (UZr)C (rapport CEA-R-3765(1)). L'addition de faibles quantites de zirconium (3 at. pour cent) au monocarbure (U,Pu)C, ameliore certaines proprietes, commee la tenue a la corrosion atmospherique, la durete et surtout la compatibilite avec l'acier inoxydable X-18 M, Par contre le coefficient de dilatation et la densite sont peu changes. Le rapport Pu/Pu+U etait fixe a 20 pour cent. Deux procedes de fabrication ont ete etudies: l'un par fusion a l'arc, l'autre par frittage a partir de metaux hydrures. Au vu des resultats metallurgiques obtenus le carbure (U,Pu,Zr)C semble presenter un interet certain. (auteur)

Techniques for chemical characterization of zirconium and its alloys

International Nuclear Information System (INIS)

Iyer, K.V.; Bassan, M.K.T.; Sudersanan, M.

2002-01-01

Chemical characterization of zirconium and its alloys such as zircaloy, Zr-Nb, etc for minor and trace constituents like Nb, Ti, Fe, Cr, Ni, Sn, Al etc has been carried out. Zirconium, being a major constituent, has been determined by gravimetry as zirconium oxide while other constituents like Nb, Ti, Fe have been determined by spectrophotometric methods. Other metals of importance at trace level have been estimated by AAS or ICPAES. The judicious use of both conventional and modern instrumental methods of analysis helps in the characterization of zirconium and its alloys for various major and minor constituents. The role of matrix effect in the determination was also investigated and methods have been worked out based on a preliminary separation of zirconium by a hydroxide precipitation. (author)

Thermal behaviour of nitrogen implanted into zirconium

International Nuclear Information System (INIS)

Miyagawa, S.; Ikeyama, M.; Saitoh, K.; Nakao, S.; Niwa, H.; Tanemura, S.; Miyagawa, Y.

1994-01-01

Zirconium films were implanted with 15 N ions of energy 50keV to a total fluence of 1x10 18 ionscm -2 in an attempt to study the formation process and thermal stability of ZrN layers produced by high fluence implantation of nitrogen. Subsequent to the implantation at room temperature, samples were annealed at temperatures of 300 C-900 C. The depth profiles of the implanted nitrogen were measured by nuclear reaction analysis using the 15 N(p,αγ) 12 C at E R =429keV, and the surfaces were examined by thin film X-ray diffraction (XRD) and scanning electron microscopy. There were many blisters 0.2-0.4μm in diameter on the surface of the as-implanted samples and double peaks were observed in the nitrogen depth profiles; they were in both sides of the mean projected range. It was found that most of the blisters became extinct after annealing above 400 C, and the XRD peak (111) intensity was increased with the increase in the annealing temperature. Moreover, 14 N and 15 N implantations were superimposed on Zr samples in order to study the atomic migration of nitrogen at each stage of high fluence implantation. It was found that the decrease in the peak at the deeper layers was related to blister extinction and nitrogen diffusion into underling zirconium which could be correlated with radiation damage induced by post-implanted ions. ((orig.))

Safe operation and maintenance of research reactor

International Nuclear Information System (INIS)

Munsorn, S.

1999-01-01

The first Thai Research Reactor (TRR-1) was established in 1961 at the Office of Atomic Energy for Peace (OAEP), Bangkok. The reactor was light water moderated and cooled, using HEU plate-type with U 3 O 8 - Al fuel meat and swimming pool type. The reactor went first critical on October 27, 1962 and had been licensed to operate at 1 MW (thermal). On June 30, 1975 the reactor was shutdown for modification and the core and control system was disassemble and replaced by that of TRIGA Mark III type while the pool cooling system, irradiation facilities and other were kept. Thus the name TRR-1/M1' has been designed due to this modification the fuel has been changed from HEU plate type to Uranium Zirconium Hydride (UZrH) Low Enrichment Uranium (LEU) which include 4 Fuel Follower Control Rods and 1 Air Follower Control Rod. The TRR-1/M1 went critical on November 7, 1977 and the purpose of the operation are training, isotope production and research. Nowadays the TRR-1/M1 has been operated with core loading No.12 which released power of 1,056 MWD. (as of October 1998). The TRR-1/M1 has been operated at the power of 1.2 MW, three days a week with 34 hours per week, Shut-down on Monday for weekly maintenance and Tuesday for special experiment. The everage energy released is about 40.8 MW-hour per week. Every year, the TRR-1/M1 is shut-down about 2 months between February to March for yearly maintenance. (author)

Safe operation and maintenance of research reactor

Energy Technology Data Exchange (ETDEWEB)

Munsorn, S. [Reactor Operation Division, Office of Atomic Energy for Peace, Chatuchak, Bangkok (Thailand)

1999-10-01

The first Thai Research Reactor (TRR-1) was established in 1961 at the Office of Atomic Energy for Peace (OAEP), Bangkok. The reactor was light water moderated and cooled, using HEU plate-type with U{sub 3}O{sub 8}- Al fuel meat and swimming pool type. The reactor went first critical on October 27, 1962 and had been licensed to operate at 1 MW (thermal). On June 30, 1975 the reactor was shutdown for modification and the core and control system was disassemble and replaced by that of TRIGA Mark III type while the pool cooling system, irradiation facilities and other were kept. Thus the name TRR-1/M1' has been designed due to this modification the fuel has been changed from HEU plate type to Uranium Zirconium Hydride (UZrH) Low Enrichment Uranium (LEU) which include 4 Fuel Follower Control Rods and 1 Air Follower Control Rod. The TRR-1/M1 went critical on November 7, 1977 and the purpose of the operation are training, isotope production and research. Nowadays the TRR-1/M1 has been operated with core loading No.12 which released power of 1,056 MWD. (as of October 1998). The TRR-1/M1 has been operated at the power of 1.2 MW, three days a week with 34 hours per week, Shut-down on Monday for weekly maintenance and Tuesday for special experiment. The everage energy released is about 40.8 MW-hour per week. Every year, the TRR-1/M1 is shut-down about 2 months between February to March for yearly maintenance. (author)

Review of the general atomic experimental fusion power reactor initial conceptual design

International Nuclear Information System (INIS)

Baker, C.C.; Sager, P.H. Jr.; Harder, C.R.

1976-01-01

The primary objective of the Experimental Power Reactor (EPR) is to provide the necessary interface between physics experiments and the first demonstration power plants. Since economically viable tokamak-type reactors may well have to be very high Q devices (ratio of fusion power out to power into the plasma), it will be essential for a tokamak demonstration reactor to operate at or near ignition conditions. Thus, it is believed that one of the primary objectives of the EPR must be to fully model the behavior of a D-T burning plasma required in the reactor of a demonstration plant. Therefore, a major objective of the EPR should be to achieve ignition conditions. In addition to demonstrating the ability to ignite and control a D-T plasma, it is also desirable that the EPR should produce, or at least demonstrate the ability to produce, a small amount of net electrical power. These objectives should be accomplished at a reasonable cost; this implies achieving a sufficiently high β (ratio of plasma pressure to magnetic field pressure). It is believed that noncircular cross section tokamaks offer the best chance of realizing these objectives. Consequently, noncircular cross sections are a major design feature of the General Atomic EPR

Fiber-optic laser-induced breakdown spectroscopy of zirconium metal in air: Special features of the plasma produced by a long-pulse laser

Science.gov (United States)

Matsumoto, Ayumu; Ohba, Hironori; Toshimitsu, Masaaki; Akaoka, Katsuaki; Ruas, Alexandre; Sakka, Tetsuo; Wakaida, Ikuo

2018-04-01

The decommissioning of the Tokyo Electric Power Company (TEPCO) Fukushima Daiichi Nuclear Power Plant is an essential issue in nuclear R&D. Fiber-optic laser-induced breakdown spectroscopy (Fiber-optic LIBS) could be used for in-situ elemental analysis of the inside of the damaged reactors. To improve the performances under difficult conditions, using a long-pulse laser can be an efficient alternative. In this work, the emission spectra of zirconium metal in air obtained for a normal-pulse laser (6 ns) and a long-pulse laser (100 ns) (wavelength: 1064 nm, pulse energy: 12.5 mJ, spot diameter: 0.35 mm) are compared to investigate the fundamental aspects of fiber-optic LIBS with the long-pulse laser. The spectral features are considerably different: when the long-pulse laser is used, the atomic and molecular emission is remarkably enhanced. The enhancement of the atomic emission at the near infrared (NIR) region would lead to the observation of emission lines with minimum overlapping. To understand the differences in the spectra induced respectively from the normal-pulse laser and the long-pulse laser, photodiode signals, time-resolved spectra, plasma parameters, emission from the ambient air, and emission regions are investigated, showing the particular characteristics of the plasma produced by the long-pulse laser.

Intercalation chemistry of zirconium 4-sulfophenylphosphonate

International Nuclear Information System (INIS)

Svoboda, Jan; Zima, Vítězslav; Melánová, Klára; Beneš, Ludvík; Trchová, Miroslava

2013-01-01

Zirconium 4-sulfophenylphosphonate is a layered material which can be employed as a host for the intercalation reactions with basic molecules. A wide range of organic compounds were chosen to represent intercalation ability of zirconium 4-sulfophenylphosphonate. These were a series of alkylamines from methylamine to dodecylamine, 1,4-phenylenediamine, p-toluidine, 1,8-diaminonaphthalene, 1-aminopyrene, imidazole, pyridine, 4,4′-bipyridine, poly(ethylene imine), and a series of amino acids from glycine to 6-aminocaproic acid. The prepared compounds were characterized by powder X-ray diffraction, thermogravimetry analysis and IR spectroscopy and probable arrangement of the guest molecules in the interlayer space of the host is proposed based on the interlayer distance of the prepared intercalates and amount of the intercalated guest molecules. - Graphical abstract: Nitrogen-containing organic compounds can be intercalated into the interlayer space of zirconium 4-sulfophenylphosphonate. - Highlights: • Zirconium 4-sulfophenylphosphonate was examined as a host material in intercalation chemistry. • A wide range of nitrogen-containing organic compounds were intercalated. • Possible arrangement of the intercalated species is described

Law project adopted by the Senate and authorizing the ratification of the additional protocol to the agreement between France, the European atomic energy community and the international atomic energy agency relative to the application of warranties in France

International Nuclear Information System (INIS)

2002-10-01

This project of law concerns an additional protocol to the agreement of warranties signed on September 22, 1998 between France, the European atomic energy community and the IAEA. This agreement concerns the declaration of all information relative to the R and D activities linked with the fuel cycle and involving the cooperation with a foreign country non endowed with nuclear weapons. These information include the trade and processing of nuclear and non-nuclear materials and equipments devoted to nuclear reactors (pressure vessels, fuel loading/unloading systems, control rods, force and zirconium tubes, primary coolant pumps, deuterium and heavy water, nuclear-grade graphite), to fuel reprocessing plants, to isotope separation plants (gaseous diffusion, laser enrichment, plasma separation, electromagnetic enrichment), to heavy water and deuterium production plants, and to uranium conversion plants. (J.S.)

Chemistry of titanium, zirconium and thorium picramates

International Nuclear Information System (INIS)

Srivastava, R.S.; Agrawal, S.P.; Bhargava, H.N.

1976-01-01

Picramates of titanium, zirconium and thorium are prepared by treating the aqueous sulphate, chloride and nitrate solutions with sodium picramate. Micro-analysis, colorimetry and spectrophotometry are used to establish the compositions (metal : ligand ratio) of these picramates as 1 : 2 (for titanium and zirconium) and 1 : 4 (for thorium). IR studies indicate H 2 N → Me coordination (where Me denotes the metal). A number of explosive properties of these picramates point to the fact that the zirconium picramate is thermally more stable than the picramates of titanium and thorium. (orig.) [de

Hydration characteristics of zirconium oxide replaced Portland cement for use as a root-end filling material.

Science.gov (United States)

Camilleri, J; Cutajar, A; Mallia, B

2011-08-01

Zirconium oxide can be added to dental materials rendering them sufficiently radiopaque. It can thus be used to replace the bismuth oxide in mineral trioxide aggregate (MTA). Replacement of Portland cement with 30% zirconium oxide mixed at a water/cement ratio of 0.3 resulted in a material with adequate physical properties. This study aimed at investigating the microstructure, pH and leaching in physiological solution of Portland cement replaced zirconium oxide at either water-powder or water-cement ratios of 0.3 for use as a root-end filling material. The hydration characteristics of the materials which exhibited optimal behavior were evaluated. Portland cement replaced by zirconium oxide in varying amounts ranging from 0 to 50% in increments of 10 was prepared and divided into two sets. One set was prepared at a constant water/cement ratio while the other set at a constant water/powder ratio of 0.3. Portland cement and MTA were used as controls. The materials were analyzed under the scanning electron microscope (SEM) and the hydration products were determined. X-ray energy dispersive analysis (EDX) was used to analyze the elemental composition of the hydration products. The pH and the amount of leachate in Hank's balanced salt solution (HBSS) were evaluated. A material that had optimal properties that satisfied set criteria and could replace MTA was selected. The microstructure of the prototype material and Portland cement used as a control was assessed after 30 days using SEM and atomic ratio diagrams of Al/Ca versus Si/Ca and S/Ca versus Al/Ca were plotted. The hydration products of Portland cement replaced with 30% zirconium oxide mixed at water/cement ratio of 0.3 were calcium silicate hydrate, calcium hydroxide and minimal amounts of ettringite and monosulphate. The calcium hydroxide leached in HBSS solution resulted in an increase in the pH value. The zirconium oxide acted as inert filler and exhibited no reaction with the hydration by-products of Portland

On the prospect of atomic engineering development in the Republic of Kazakhstan

International Nuclear Information System (INIS)

Batyrbekov, G.A.; Mahanov, U. M.

2002-01-01

: classical way using of light water reactors and imported enriched uranium, using of heavy water reactors on natural uranium of Candu type, using a fast neutron reactors NPP construction. The advantages and lack of each way in Kazakhstan are analysed here. Different combination of these ways is possible for the first stage of works. Production of zirconium metal and zirconium rolled product for manufacturing of zirconium shell for fuel elements in Kazakhstan will be provided by using of the Obukhovsk titanium-zirconium ore deposit (near Kokshetau town) for any of the above mentioned ways. It is possible to organize a long storage of used fuel like state storehouse in Sweden while the problem a different burial ways will be solve

Hydride formation on deformation twin in zirconium alloy

Energy Technology Data Exchange (ETDEWEB)

Kim, Ju-Seong [Korea Atomic Energy Research Institute, 989-111 Daedeokdaero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of); Kim, Sung-Dae [Korea Institute of Material Science (KIMS), 797 Changwondaero, Changwon, Gyeongnam, 642-831 (Korea, Republic of); Yoon, Jonghun, E-mail: yooncsmd@gmail.com [Department of Mechanical Engineering, Hanyang University, 1271 Sa3-dong, Sangrok-gu, Ansan-si, Gyeonggi-do, 426-791 (Korea, Republic of)

2016-12-15

Hydrides deteriorate the mechanical properties of zirconium (Zr) alloys used in nuclear reactors. Intergranular hydrides that form along grain boundaries have been extensively studied due to their detrimental effects on cracking. However, it has been little concerns on formation of Zr hydrides correlated with deformation twins which is distinctive heterogeneous nucleation site in hexagonal close-packed metals. In this paper, the heterogeneous precipitation of Zr hydrides at the twin boundaries was visualized using transmission electron microscopy. It demonstrates that intragranular hydrides in the twinned region precipitates on the rotated habit plane by the twinning and intergranular hydrides precipitate along the coherent low energy twin boundaries independent of the conventional habit planes. Interestingly, dislocations around the twin boundaries play a substantial role in the nucleation of Zr hydrides by reducing the misfit strain energy.

21 CFR 700.16 - Use of aerosol cosmetic products containing zirconium.

Science.gov (United States)

2010-04-01

... zirconium. 700.16 Section 700.16 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN... cosmetic products containing zirconium. (a) Zirconium-containing complexes have been used as an ingredient... indicates that certain zirconium compounds have caused human skin granulomas and toxic effects in the lungs...

Report on application results of the nuclear reactor in Atomic Energy Research Laboratory, Rikkyo University. April 1994 - March 1995

International Nuclear Information System (INIS)

1995-01-01

This report is on researching action state, application state, management state, and others of 1994 fiscal year at the Atomic Energy Research Laboratory, Rikkyo University. The experimental reactor has been used for the studies such as application of neutron radioactivity analysis to multi fields, application of fission and alpha track method to age determination and metallurgy, hot atom chemistry, neutron radiation effect on semiconductors and others, nuclear data measurement, organism, materials and products using neutron radiography, and development and application to inspection of radiation detectors such as neutron detector. This report was a report shown as a shape of research results of actions of the researchers. And, another report of colaborate research results using the Rikkyo University reactor was also published from the Atomic Energy Center, the University of Tokyo begun since April, 1974. (G.K.)

Processing fissile material mixtures containing zirconium and/or carbon

Science.gov (United States)

Johnson, Michael Ernest; Maloney, Martin David

2013-07-02

A method of processing spent TRIZO-coated nuclear fuel may include adding fluoride to complex zirconium present in a dissolved TRIZO-coated fuel. Complexing the zirconium with fluoride may reduce or eliminate the potential for zirconium to interfere with the extraction of uranium and/or transuranics from fission materials in the spent nuclear fuel.

Hot zirconium cathode sputtered layers for useful surface modification

International Nuclear Information System (INIS)

Duckworth, R.G.

1986-01-01

It has been found that multilayer zirconium based sputtered coatings can greatly improve the wear properties of a wide variety of mechanical components, machine tools, and metal surfaces. Although a hot (approximately 1000 0 C) cathode is employed, temperature sensitive components can be beneficially treated, and for precision parts a total coating thickness of only 0.5μm is often perfectly effective. Even at the highest coating rates substrate temperatures are below 300 0 C. For the corrosion protection of less well finished surfaces thicker layers are usually required and it is important that relatively stress free layers are produced. The authors employed a variety of tailored zirconium/zirconium nitride/zirconium oxide mixed layers to solve a number of tribological problems for some 5 or 6 years. However, it is only recently that they designed, built, and commissioned rapid cycle, multiple cathode, load-lock plant for economic production of such coatings. This paper provides an introduction to this method of depositing pure zirconium and pure synthetic zirconium nitride films

Determination of microquantities of zirconium and thorium in uranium dioxide

International Nuclear Information System (INIS)

Weber de D'Alessio, Ana; Zucal, Raquel.

1975-07-01

A method for the determination of 10 to 50 ppm of zirconium and thorium in uranium IV oxide of nuclear purity is established. Zirconium and thorium are retained in a strong cation-exchange resin Dowex 50 WX8 in 1 M HCl. Zirconium is eluted with 0,5% oxalic acid solution and thorium with 4% ammonium oxalate. The colorimetric determination of zirconium with xilenol orange is done in perchloric acid after destructtion of oxalic acid and thorium is determined with arsenazo III in 5 M HCl. 10 μg of each element were determined with a standard deviation of 2,1% for thorium and 3,4% for zirconium. (author) [es

Radiation Damage and Fission Product Release in Zirconium Nitride

Energy Technology Data Exchange (ETDEWEB)

Egeland, Gerald W. [New Mexico Inst. of Mining and Technology, Socorro, NM (United States)

2005-08-29

Zirconium nitride is a material of interest to the AFCI program due to some of its particular properties, such as its high melting point, strength and thermal conductivity. It is to be used as an inert matrix or diluent with a nuclear fuel based on transuranics. As such, it must sustain not only high temperatures, but also continuous irradiation from fission and decay products. This study addresses the issues of irradiation damage and fission product retention in zirconium nitride through an assessment of defects that are produced, how they react, and how predictions can be made as to the overall lifespan of the complete nuclear fuel package. Ion irradiation experiments are a standard method for producing radiation damage to a surface for observation. Cryogenic irradiations are performed to produce the maximum accumulation of defects, while elevated temperature irradiations may be used to allow defects to migrate and react to form clusters and loops. Cross-sectional transmission electron microscopy and grazing-incidence x-ray diffractometry were used in evaluating the effects that irradiation has on the crystal structure and microstructure of the material. Other techniques were employed to evaluate physical effects, such as nanoindentation and helium release measurements. Results of the irradiations showed that, at cryogenic temperatures, ZrN withstood over 200 displacements per atom without amorphization. No significant change to the lattice or microstructure was observed. At elevated temperatures, the large amount of damage showed mobility, but did not anneal significantly. Defect clustering was possibly observed, yet the size was too small to evaluate, and bubble formation was not observed. Defects, specifically nitrogen vacancies, affect the mechanical behavior of ZrN dramatically. Current and previous work on dislocations shows a distinct change in slip plane, which is evidence of the bonding characteristics. The stacking-fault energy changes dramatically with

Design and analysis of reactor headers for Narora Atomic Power Project

International Nuclear Information System (INIS)

Danak, M.R.

1975-01-01

Reactor header for Narora Atomic Power Reactor is a 400 mm O.D. 10 metres long pressure vessel in the primary coolant circuit connecting 153 feeders to PHT pumps or steam generators. The vessel dimensions are restricted are by containment philosophy. The outlet connections for pumps or steam generators are to be of the size of vessel diameter and DO/t ratio for the vessel is approximately 10. The design and stresses induced meet the code requirements except that at times it is difficult to get precise stress values in absence of certain data and lack of code or available literature giving practical approach to the problem. It can be seen that the 400 mm equal tees used as part of the vessel cannot be penetrated in the light of code reinforcement requirements. However if the tees have to penetrated to retain established feeder layout, it should be established experimentally or by some detailed stress analysis that it will meet the intent of code. (author)

Analysis of hydrogen in zirconium metallic

International Nuclear Information System (INIS)

Rodrigues, A.N.; Vega Bustillos, J.O.W.

1991-02-01

Determination of hydrogen in zirconium metallic have been performed using the hot vacuum extraction system and the gas chromatographic technique. The zirconium metallic samples were hydrieded by electrolitic technique at difference temperatures and times, then the samples were annealing at vacuum and eatching by fluoridric acid solution. The details of the hydrieded process, analytical technique and the data obtained are discussed. (author)

Delayed Hydride Cracking Mechanism in Zirconium Alloys and Technical Requirements for In-Service Evaluation of Zr-2.5Nb Tubes with Flaws

International Nuclear Information System (INIS)

Kim, Young Suk

2007-01-01

In association with periodic inspection of CANDU nuclear power plant components, Canadian Standards Association issued CSA N285.8 in 2005 as technical requirements for in-service evaluation of zirconium alloy pressure tubes in CANDU reactors. This first version, CSA N285.8 involves procedures for, firstly, the evaluation of pressure tube flaws, secondly, the evaluation of pressure tube to calandria tube contact and, thirdly, the assessment of a reactor core, and material properties and derived quantities. The evaluation of pressure tube flaws includes delayed hydride cracking evaluation the procedures of which are stipulated based on the existing delayed hydride cracking models. For example, the evaluation of flaw-tip hydride precipitation during reactor cooldown involves a procedure to calculate the equilibrium hydrogen equivalent concentration in solution at the flaw tip, Htipas follows: Htip=Hfexp[- (VH delta no.)/RT], where Hf is the total bulk hydrogen equivalent concentration, VH partial molar volume of hydrogen in zirconium, δ a difference in hydrostatic stress between the bulk and the crack tip. When Htip ≥TSSP at temperature, then flaw-tip hydride is predicted to precipitate. Eq. (1) suggests that hydrogen concentration at the crack tip would increase due to an work energy given by the difference in the hydrostatic stress

Separation process of zirconium and hafnium; Procede de separation du zirconium et du hafnium

Energy Technology Data Exchange (ETDEWEB)

Hure, J; Saint-James, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1955-07-01

About the separation different processes of zirconium-hafnium, the extraction by solvent in cross-current is the most easily the process usable on an industrial scale. It uses tributyl phosphate as solvent, diluted with white spirit to facilitate the decanting. Some exploratory tests showed that nitric environment seemed the most favorable for extraction; but a lot of other factors intervene in the separation process. We studied the influence of the acidity successively, the NO{sub 3}{sup -} ions concentration, the role of the cation coming with NO{sub 3}{sup -}, as well as the influence of the concentration of zirconium in the solution on the separation coefficient {beta} = {alpha}{sub Zr} / {alpha}{sub Hf}. (M.B.) [French] Des differents procedes de separation zirconium-hafnium, l'extraction par solvant en contre-courant est le procede le plus facilement utilisable a l'echelle industrielle. On utilise comme solvant le phosphate de tributyle, dilue avec du white spirit pour faciliter les decantations. Des essais preliminaires ont montre que le milieu nitrique semblait le plus favorable a l'extraction; mais beaucoup d'autres facteurs interviennent dans le processus de separation. Nous avons etudie successivement l'influence de l'acidite, celle de la concentration en ions NO{sub 3}{sup -}, le role du cation accompagnant NO{sub 3}{sup -}, ainsi que l'influence de la concentration en zirconium de la solution sur le coefficient de separation {beta} = {alpha}{sub Zr} / {alpha}{sub Hf}. (MB)

Experimental approach and micro-mechanical modeling of the mechanical behavior of irradiated zirconium alloys

International Nuclear Information System (INIS)

Onimus, F.

2003-12-01

Zirconium alloys cladding tubes containing nuclear fuel of the Pressurized Water Reactors constitute the first safety barrier against the dissemination of radioactive elements. Thus, it is essential to predict the mechanical behavior of the material in-reactor conditions. This study aims, on the one hand, to identify and characterize the mechanisms of the plastic deformation of irradiated zirconium alloys and, on the other hand, to propose a micro-mechanical modeling based on these mechanisms. The experimental analysis shows that, for the irradiated material, the plastic deformation occurs by dislocation channeling. For transverse tensile test and internal pressure test this channeling occurs in the basal planes. However, for axial tensile test, the study revealed that the plastic deformation also occurs by channeling but in the prismatic and pyramidal planes. In addition, the study of the macroscopic mechanical behavior, compared to the deformation mechanisms observed by TEM, suggested that the internal stress is higher in the case of irradiated material than in the case of non-irradiated material, because of the very heterogeneous character of the plastic deformation. This analysis led to a coherent interpretation of the mechanical behavior of irradiated materials, in terms of deformation mechanisms. The mechanical behavior of irradiated materials was finally modeled by applying homogenization methods for heterogeneous materials. This model is able to reproduce adequately the mechanical behavior of the irradiated material, in agreement with the TEM observations. (author)

Quantum Mechanical Calculations Of Elastic Properties Of Doped Tetragonal Yttria-Stabilized Zirconium Dioxide

Directory of Open Access Journals (Sweden)

Yuriy Natanzon

2008-01-01

Full Text Available We report first principles calculations of the electronic and elastic properties of yttriastabilized tetragonal zirconium dioxide doped with metal oxides like: GeO2, TiO2, SiO2,MgO and Al2O3. It is shown that addition of such dopants affects selected elastic propertiesof ZrO2, which is driven by the attraction of electron density by dopant atom and creationof stronger dopant–oxygen bonds. This effect contributes to the increase of superplasticityof doped material.

A new mechanism of hydrogen absorption in water-water reactor core materials

International Nuclear Information System (INIS)

Gann, V.V.; Gann, A.V.

2012-01-01

The spectrum of fast protons, generated in water by fast neutrons of WWER-1000 reactor core, has been calculated using the code MCNPX. The main mechanism of fast proton generation in the moderator is found to be elastic scattering of fast neutrons on hydrogen nuclei. Fast protons with mean energy 1 MeV flow towards the surface of cladding material at flux density ∼ 0.1 μA/cm 2 . Proton range distribution profile in cladding material is calculated. The range of fast protons in zirconium averages 20 μm, the maximal proton range is larger than 200 μm. The rate of hydrogen deposition in 40 μm layer amounts to 5 x 10 -5 H/n/μ. A role of the suggested mechanism in process of zirconium clad hydrogenation during reactor irradiation is discussed.

A computation model for the corrosion resistance of nanocrystalline zirconium metal

International Nuclear Information System (INIS)

Zhang Xiyan; Shi Minghua; Liu Nianfu; Wei Yiming; Li Cong; Qiu Shaoyu; Zhang Qiang; Zhang Pengcheng

2007-01-01

In this paper a computation model of corrosion rate-grain size of nanocrystalline and ultra-fine zirconium has been presented. The model is based on the Wagner's theory and the electron theory of solids. The conductivity, electronic mean free path and grain size of metal were considered. By this model, the corrosion rate of zirconium metal under different temperature was computed. The results show that the corrosion weight gain and rate constant of nanocrystalline zirconium is lower than that of zirconium with coarse grain size. And the corrosion rate constant and weight gain of nanocrystalline zirconium metal decrease with the decrease of grain size. So the refinement of grain size can remarkably improve the corrosion resistance of zirconium metal. (authors)

In situ monitored in-pile creep testing of zirconium alloys

Science.gov (United States)

Kozar, R. W.; Jaworski, A. W.; Webb, T. W.; Smith, R. W.

2014-01-01

The experiments described herein were designed to investigate the detailed irradiation creep behavior of zirconium based alloys in the HALDEN Reactor spectrum. The HALDEN Test Reactor has the unique capability to control both applied stress and temperature independently and externally for each specimen while the specimen is in-reactor and under fast neutron flux. The ability to monitor in situ the creep rates following a stress and temperature change made possible the characterization of creep behavior over a wide stress-strain-rate-temperature design space for two model experimental heats, Zircaloy-2 and Zircaloy-2 + 1 wt%Nb, with only 12 test specimens in a 100-day in-pile creep test program. Zircaloy-2 specimens with and without 1 wt% Nb additions were tested at irradiation temperatures of 561 K and 616 K and stresses ranging from 69 MPa to 455 MPa. Various steady state creep models were evaluated against the experimental results. The irradiation creep model proposed by Nichols that separates creep behavior into low, intermediate, and high stress regimes was the best model for predicting steady-state creep rates. Dislocation-based primary creep, rather than diffusion-based transient irradiation creep, was identified as the mechanism controlling deformation during the transitional period of evolving creep rate following a step change to different test conditions.

Electrochemical stripping determination of traces of copper, lead, cadmium and zinc in zirconium metal and zirconium dioxide

International Nuclear Information System (INIS)

Stulik, K.; Beran, P.; Dolezal, J.; Opekar, F.

1978-01-01

Procedures have been developed for the determination of copper, lead, cadmium and zinc in zirconium metal and zirconium dioxide, at concentrations of 1ppm or less. Zirconium metal was dissolved in sulphuric acid, and zirconium dioxide decomposed under pressure with hydrofluoric acid. Sample solutions were prepared in dilute sulphuric acid. For the stripping determination, the sample solution was either mixed with a complexing tartrate base electrolyte or the pre-electrolysis was carried out in acid solution, with the acid solution being exchanged for a pure base electrolyte (e.g. an acetate buffer) for the stripping step. The stripping step was monitored by d.c., differential pulse and Kalousek commutator voltammetry and the three methods were compared. A stationary mercury-drop electrode can generally be used for all the methods, whereas a mercury-film electrode is suitable only for the d.c. voltammetric determination of copper, lead and cadmium, as pulse measurements with films are poorly reproducible and the electrodes are easily damaged. The relative standard deviation does not exceed 20%. Some samples contained relatively large amounts of copper, which is best separated by electrodeposition on a platinum electrode. (author)

Preliminary irradiation test results from the Yankee Atomic Electric Company reactor vessel test irradiation program

International Nuclear Information System (INIS)

Biemiller, E.C.; Fyfitch, S.; Campbell, C.A.

1993-01-01

The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. Plate materials each containing 0.24 w/o copper, but different nickel contents at 0.63 w/o and 0.19 w/o, were heat treated to simulate the Yankee vessel heat treatment (austenitized at 1800 deg F) and to simulate Regulatory Guide 1.99 database materials (austenitized at 1600 deg. F). These heat treatments produced different microstructures so the effect of microstructure on irradiation damage sensitivity could be tested. Because the nickel content of the test plates varied and the copper level was constant, the effect of nickel on irradiation embrittlement was also tested. Correlation monitor material, HSST-02, was included in the program to benchmark the Ford Nuclear Reactor (U. of Michigan Test Reactor) which had never been used for this type of irradiation program. Materials taken from plate surface locations (vs. 1/4T) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from the rapid quench, is maintained after irradiation. If the improved properties are maintained, pressurized thermal shock calculations could utilize this margin. Finally, for one experiment, irradiations were conducted at two irradiation temperatures (500 deg. F and 550 deg. F) to determine the effect of irradiation temperature on embrittlement. The preliminary results of the irradiation program show an increase in T 30 shift of 69 deg. F for a decrease in irradiation temperature of 50 deg. F. The results suggest that for nickel bearing steels, the superior toughness of plate surface material is maintained after irradiation and for the copper content tested, nickel had no apparent effect on irradiation response. No apparent microstructure

Effect of self-interstitial diffusion anisotropy in electron-irradiated zirconium: A cluster dynamics modeling

International Nuclear Information System (INIS)

Christien, F.; Barbu, A.

2005-01-01

A model based on the cluster dynamics approach was proposed in [A. Hardouin Duparc, C. Moingeon, N. Smetniansky-de-Grande, A. Barbu, J. Nucl. Mater. 302 (2002) 143] to describe point defect agglomeration in metals under irradiation. This model is restricted to materials where point defect diffusion is isotropic and is thus not applicable to anisotropic metals such as zirconium. Following the approach proposed by Woo [C.H. Woo, J. Nucl. Mater. 159 (1988) 237], we extended in this work the model to the case where self-interstitial atoms (SIA) diffusion is anisotropic. The model was then applied to the loop microstructure evolution of a zirconium thin foil irradiated with electrons in a high-voltage microscope. First, the inputs were validated by comparing the numerical results with Hellio et al. experimental results [C. Hellio, C.H. de Novion, L. Boulanger, J. Nucl. Mater. 159 (1988) 368]. Further calculations were made to evidence the effect of the thin foil orientation on the dislocation loop microstructure under irradiation. The result is that it is possible to reproduce for certain orientations the 'unexpected' vacancy loop growth experimentally observed in electron-irradiated zirconium [M. Griffiths, M.H. Loretto, R.E. Sallmann, J. Nucl. Mater. 115 (1983) 313; J. Nucl. Mater. 115 (1983) 323; Philos. Mag. A 49 (1984) 613]. This effect is directly linked to SIA diffusion anisotropy

40 CFR 721.10089 - Modified salicylic acid, zirconium complex (generic).

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 30 2010-07-01 2010-07-01 false Modified salicylic acid, zirconium... Specific Chemical Substances § 721.10089 Modified salicylic acid, zirconium complex (generic). (a) Chemical... as modified salicylic acid, zirconium complex (PMN P-00-552) is subject to reporting under this...

Preparation of complexes of zirconium and hafnium tetrachlorides with phosphorus oxychloride

International Nuclear Information System (INIS)

McLaughlin, D.F.

1989-01-01

This patent describes an improvement in a method for separating hafnium tetrachloride from zirconium tetrachloride where a complex of zirconium-hafnium tetrachlorides and phosphorus oxychloride is prepared from zirconium-hafnium tetrachlorides and the complex of zirconium-hafnium tetrachlorides and phosphorus oxychloride is introduced into a distillation column, which distillation column has a reboiler connected at the bottom and a reflux condenser connected at the top and where a hafnium tetrachloride enriched stream is taken from the top of the column and a zirconium enriched tetrachloride stream is taken from the bottom of the column. The improvement comprising: prepurifying the zirconium-hafnium tetrachlorides, prior to preparation of the complex and introduction of the complex into a distillation column, to substantially eliminate iron chloride from the zirconium hafnium tetrachlorides, whereby buildup or iron chloride in the distillation column and in the reboiler is substantially eliminated and the column can be operated in a continuous, stable and efficient manner

Low cycle fatigue behaviour of zirconium alloys at 3000C

International Nuclear Information System (INIS)

Hosbons, R.R.

1975-01-01

The low cycle fatigue lives of two zirconium alloys, zirconium--2.5 wt percent niobium and zirconium--1.1 wt percent chromium--0.1 wt percent iron, have been determined at 300 0 C. Both annealed material and cold-worked and stress-relieved material have similar fatigue lives to annealed Zircaloy-2 but β-quenched zirconium--niobium and zirconium--chromium--iron have lower fatigue lives than annealed Zircaloy-2. An atmosphere containing a concentration of iodine lower than that required for stress corrosion cracking still significantly lowers the fatigue life. A mathematical relationship between fatigue life and short-term tensile properties was used to estimate the fatigue life of zirconium alloy fuel sheaths and it was estimated that for a strain cycle of 0.1 percent a cyclic frequency exceeding 0.116 Hz (10,000 cycles/ day) would be required to cause fatigue failure of the sheath before its design life is realized

Low cycle fatigue behaviour of zirconium alloys at 3000C

International Nuclear Information System (INIS)

Hosbons, R.R.

1975-01-01

The low cycle fatigue lives of two zirconium alloys, zirconium-2.5 wt% niobium and zirconium-1.1 wt% chronium-0.1 wt% iron, have been determined at 300 0 C. Both annealed material and cold-worked and stress-relieved material have similar fatigue lives to annealed Zircaloy-2 but β-quenched zirconium-niobium and zirconium-chromium-iron have lower fatigue lives than annealed Zircaloy-2. An atmosphere containing a concentration of iodine lower than that required for stress corrosion cracking still significantly lowers the fatigue life. A mathematical relationship between fatigue life and short-term tensile properties was used to estimate the fatigue life of zirconium alloy fuel sheaths and it was estimated that for a strain cycle of 0.1 per cent a cyclic frequency exceeding 0.116 Hz (10 000 cycles/day) would be required to cause fatigue failure of the sheath before its design life is realized. (author)

Investigation of anodic oxide coatings on zirconium after heat treatment

International Nuclear Information System (INIS)

Sowa, Maciej; Dercz, Grzegorz; Suchanek, Katarzyna; Simka, Wojciech

2015-01-01

Highlights: • Oxide layers prepared via PEO of zirconium were subjected to heat treatment. • Surface characteristics were determined for the obtained oxide coatings. • Heat treatment led to the partial destruction of the anodic oxide layer. • Pitting corrosion resistance of zirconium was improved after the modification. - Abstract: Herein, results of heat treatment of zirconium anodised under plasma electrolytic oxidation (PEO) conditions at 500–800 °C are presented. The obtained oxide films were investigated by means of SEM, XRD and Raman spectroscopy. The corrosion resistance of the zirconium specimens was evaluated in Ringer's solution. A bilayer oxide coatings generated in the course of PEO of zirconium were not observed after the heat treatment. The resulting oxide layers contained a new sublayer located at the metal/oxide interface is suggested to originate from the thermal oxidation of zirconium. The corrosion resistance of the anodised metal was improved after the heat treatment

Rotary reactor for atomic layer deposition on large quantities of nanoparticles

International Nuclear Information System (INIS)

McCormick, J. A.; Cloutier, B. L.; Weimer, A. W.; George, S. M.

2007-01-01

Challenges are encountered during atomic layer deposition (ALD) on large quantities of nanoparticles. The particles must be agitated or fluidized to perform the ALD surface reactions in reasonable times and to prevent the particles from being agglomerated by the ALD film. The high surface area of nanoparticles also demands efficient reactant usage because large quantities of reactant are required for the surface reactions to reach completion. The residence time of the reactant in a fluidized particle bed reactor may be too short for high efficiency if the ALD surface reactions have low reactive sticking coefficients. To address these challenges, a novel rotary reactor was developed to achieve constant particle agitation during static ALD reactant exposures. In the design of this new reactor, a cylindrical drum with porous metal walls was positioned inside a vacuum chamber. The porous cylindrical drum was rotated by a magnetically coupled rotary feedthrough. By rotating the cylindrical drum to obtain a centrifugal force of less than one gravitational force, the particles were agitated by a continuous 'avalanche' of particles. In addition, an inert N 2 gas pulse helped to dislodge the particles from the porous walls and provided an efficient method to purge reactants and products from the particle bed. The effectiveness of this rotary reactor was demonstrated by Al 2 O 3 ALD on ZrO 2 particles. A number of techniques including transmission electron microscopy, Fourier transform infrared spectroscopy, and x-ray photoelectron spectroscopy confirmed that the Al 2 O 3 ALD film conformally coats the ZrO 2 particles. Combining static reactant exposures with a very high surface area sample in the rotary reactor also provides unique opportunities for studying the surface chemistry during ALD

Simulating the behaviour of zirconium-alloy components in nuclear reactors

International Nuclear Information System (INIS)

Coleman, C.E.

2001-12-01

To prevent failure in nuclear components one needs to understand the interactions between adjacent materials and the changes in their physical properties during all phases of reactor operation. Three examples from CANDU reactors are described to illustrate the use of simulations that imitate complicated reactor situations. These are: swelling tests that led to a method for increasing the tolerance or Zircaloy fuel cladding to power ramps; observations of the behaviour of leaking cracks in Zr-2.5Nb pressure tubes that provide confidence in the use of leak-before-break as part of the defence against flaw development; and contact boiling tests on modifications to the surfaces of Zircaloy calandria tubes that enhance the ability of the heavy water moderator to act as a heat sink after a postulated loss-of-coolant accident. (author)

Manufacture of fuel and fuel channels and their performance in Indian PHWRs'

International Nuclear Information System (INIS)

Kalidas, R.

2005-01-01

Nuclear Fuel Complex (NFC) at Hyderabad is conglomeration of chemical, metallurgical and mechanical plants, processing uranium and zirconium in two separate streams and culminating in the fuel assembly plant. Apart from manufacturing fuel for Pressurised Heavy Water Reactors (PHWRs) and Boiling Water Reactors (BWRs), NFC is also engaged in the manufacture of reactor core structurals for these reactors. NFC has carried our several technological developments over the years and implemented them for the manufacture of fuel, calandria tubes and pressure tubes for PHWRs. Keeping in pace with the Nuclear Power Programme envisaged by the Department of Atomic Energy, NFC had augmented its production capacities in all these areas. The paper highlights several actions initiated in the areas of fuel design, fuel manufacturing, manufacturing of zirconium alloy core structurals, fuel clad tubes and components and their performance in Indian PHWRs. (author)

Manufacture of fuel and fuel channels and their performance in Indian PHWRS - an overview

International Nuclear Information System (INIS)

Kalidas, R.

2005-01-01

Nuclear Fuel Complex (NFC) at Hyderabad is a conglomeration of chemical, metallurgical and mechanical plants, processing uranium and zirconium in two separate streams and culminating in the fuel assembly plant. Apart from manufacturing fuel for Pressurised Heavy Water Reactors (PHWRs) and Boiling Water Reactors (BWRs), NFC is also engaged in the manufacture of reactor core structurals for these reactors. NFC has carried out several technological developments over the years and implemented them for the manufacture of fuel, calandria tubes and pressure tubes for PHWRs. Keeping in pace with the Nuclear Power Programme envisaged by the Department of Atomic Energy, NFC had augmented its production capacities in all these areas. The paper highlights several actions initiated in the areas of fuel design, fuel manufacturing, manufacturing of zirconium alloy core structurals, fuel clad tubes and components and their performance in Indian PHWRs. (author)

Understanding the Irradiation Behavior of Zirconium Carbide

International Nuclear Information System (INIS)

Motta, Arthur; Sridharan, Kumar; Morgan, Dane; Szlufarska, Izabela

2013-01-01

Zirconium carbide (ZrC) is being considered for utilization in high-temperature gas-cooled reactor fuels in deep-burn TRISO fuel. Zirconium carbide possesses a cubic B1-type crystal structure with a high melting point, exceptional hardness, and good thermal and electrical conductivities. The use of ZrC as part of the TRISO fuel requires a thorough understanding of its irradiation response. However, the radiation effects on ZrC are still poorly understood. The majority of the existing research is focused on the radiation damage phenomena at higher temperatures (>450ee)C) where many fundamental aspects of defect production and kinetics cannot be easily distinguished. Little is known about basic defect formation, clustering, and evolution of ZrC under irradiation, although some atomistic simulation and phenomenological studies have been performed. Such detailed information is needed to construct a model describing the microstructural evolution in fast-neutron irradiated materials that will be of great technological importance for the development of ZrC-based fuel. The goal of the proposed project is to gain fundamental understanding of the radiation-induced defect formation in zirconium carbide and irradiation response by using a combination of state-of-the-art experimental methods and atomistic modeling. This project will combine (1) in situ ion irradiation at a specialized facility at a national laboratory, (2) controlled temperature proton irradiation on bulk samples, and (3) atomistic modeling to gain a fundamental understanding of defect formation in ZrC. The proposed project will cover the irradiation temperatures from cryogenic temperature to as high as 800ee)C, and dose ranges from 0.1 to 100 dpa. The examination of this wide range of temperatures and doses allows us to obtain an experimental data set that can be effectively used to exercise and benchmark the computer calculations of defect properties. Combining the examination of radiation

RECOVERY OF URANIUM FROM ZIRCONIUM-URANIUM NUCLEAR FUELS

Science.gov (United States)

Gens, T.A.

1962-07-10

An improvement was made in a process of recovering uranium from a uranium-zirconium composition which was hydrochlorinated with gsseous hydrogen chloride at a temperature of from 350 to 800 deg C resulting in volatilization of the zirconium, as zirconium tetrachloride, and the formation of a uranium containing nitric acid insoluble residue. The improvement consists of reacting the nitric acid insoluble hydrochlorination residue with gaseous carbon tetrachloride at a temperature in the range 550 to 600 deg C, and thereafter recovering the resulting uranium chloride vapors. (AEC)

Composition and Performance of Nanostructured Zirconium Titanium Conversion Coating on Aluminum-Magnesium Alloys

Directory of Open Access Journals (Sweden)

Sheng-xue Yu

2013-01-01

Full Text Available Nanostructured conversion coating of Al-Mg alloy was obtained via the surface treatment with zirconium titanium salt solution at 25°C for 10 min. The zirconium titanium salt solution is composed of tannic acid 1.00 g·L−1, K2ZrF6 0.75 g·L−1, NaF 1.25 g·L−1, MgSO4 1.0 g/L, and tetra-n-butyl titanate (TBT 0.08 g·L−1. X-ray diffraction (XRD, X-ray photoelectron spectroscopy (XPS, and Fourier transform infrared spectrum (FT-IR were used to characterize the composition and structure of the obtained conversion coating. The morphology of the conversion coating was obtained by atomic force microscopy (AFM and scanning electron microscopy (SEM. Results exhibit that the zirconium titanium salt conversion coating of Al-Mg alloy contains Ti, Zr, Al, F, O, Mg, C, Na, and so on. The conversion coating with nm level thickness is smooth, uniform, and compact. Corrosion resistance of conversion coating was evaluated in the 3.5 wt.% NaCl electrolyte through polarization curves and electrochemical impedance spectrum (EIS. Self-corrosion current density on the nanostructured conversion coating of Al-Mg alloy is 9.7×10-8A·cm-2, which is only 2% of that on the untreated aluminum-magnesium alloy. This result indicates that the corrosion resistance of the conversion coating is improved markedly after chemical conversion treatment.

Thai research reactor

International Nuclear Information System (INIS)

Aramrattana, M.

1987-01-01

The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)

Mechanisms of irradiation growth of alpha-zirconium alloys

International Nuclear Information System (INIS)

Holt, R.A.

1988-01-01

Experimental observations in the last few years have shown that the range of irradiation growth behaviour of alpha-zirconium alloys is more varied, that a wider variety of sinks must be considered, and that there are more potential sources of anisotropy than was previously recognized. The important new experimental observations which influence our preception of the growth phenomenon in zirconium alloys include the growth of single crystals, accelerating growth in annealed material with the coincident appearance of vacancy loops on the basal planes, the occurrence of 'negative' growth, i.e., contractions along prism directions, the absence of a pronounced effect of grain size on the long term growth rate at low temperatures, and the presence of intergranular constraints prior to irradiation. With the greater complexity of behaviour now being observed, it is necessary to apply new theoretical concepts to assist in understanding growth, e.g., the potential role of anisotropic diffusion in segregation point defects to different sinks and 'growth' caused by the anisotropic relaxation of intergranular constrains. These can be combined with earlier ideas to predict a variety of growth behaviours, including 'negative growth'. Because the most important physical information required for theoretical treatments of growth, i.e, the characteristics of vacancies and self interstitial atoms, are still poorly understood, it is almost impossible to test rigorously any particular theoretical concept and a complete picture of growth has yet to emerge. (orig./MM)

PROCESS OF DISSOLVING ZIRCONIUM ALLOYS

Science.gov (United States)

Shor, R.S.; Vogler, S.

1958-01-21

A process is described for dissolving binary zirconium-uranium alloys where the uranium content is about 2%. In prior dissolution procedures for these alloys, an oxidizing agent was added to prevent the precipitation of uranium tetrafluoride. In the present method complete dissolution is accomplished without the use of the oxidizing agent by using only the stoichiometric amount or slight excess of HF required by the zirconium. The concentration of the acid may range from 2M to 10M and the dissolution is advatageously carried out at a temperature of 80 deg C.

Intercomparison of techniques for inspection and diagnostics of heavy water reactor pressure tubes: Flaw detection and characterization [Phase 1

International Nuclear Information System (INIS)

2006-05-01

Nuclear power plants with heavy water reactors (HWRs) comprise nine percent of today's operating nuclear units, and more are under construction. Efficient and accurate inspection and diagnostic techniques for various reactor components and systems are an important factor in assuring reliable and safe plant operation. To foster international collaboration in the efficient and safe use of nuclear power, the IAEA conducted a Coordinated Research Programme (CRP) on Inter-comparison of Techniques for HWR Pressure Tube Inspection and Diagnostics. This CRP was carried out within the frame of the IAEA Department of Nuclear Energy's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). The TWG-HWR is a group of experts nominated by their governments and designated by the IAEA to provide advice and to support implementation of the IAEA's project on advanced technologies for HWRs. The objective of the CRP was to inter-compare non-destructive inspection and diagnostic techniques, in use and being developed, for structural integrity assessment of HWR pressure tubes. During the first phase of this CRP, participants have investigated the capability of different techniques to detect and characterize flaws. During the second phase of this CRP, participants collaborated to detect and characterize hydride blisters and to determine the hydrogen concentration in Zirconium alloys. The intent was to identify the most effective pressure tube inspection and diagnostic methods, and to identify further development needs. The organizations that have participated in this CRP are: - The Comision Nacional de Energia Atomica (CNEA), Argentina; - Atomic Energy of Canada Ltd. (AECL); Chalk River Laboratories (CRL), Canada; - The Research Institute of Nuclear Power Operations (RINPO), China National Nuclear Corporation (CNNC), China; - Bhabha Atomic Research Centre (BARC), India; - The Korea Electric Power Research Institute (KEPRI), Republic of Korea; - The Korea Atomic Energy

Characterisation of hydrides in a zirconium alloy, by EBSD

International Nuclear Information System (INIS)

Ubhi, H.S.; Larsen, K.

2012-01-01

Zirconium alloys are used in nuclear reactors owing to their low capture cross-section for thermal neutrons and good mechanical and corrosion properties. However, they do suffer from delayed hydrogen cracking (DHC) due to formation of hydride particles. This study shows how the electron back-scatter diffraction (EBSD) technique can be used to characterise hydrides and their orientation relationship with the matrix. Hydrided EB weld specimens were prepared by electro-polishing, characterised using Oxford instruments AZtecHKL EBSD apparatus and software attached to a FEG SEM. Hydrides were found to exist as fine intra granular plates and having the Blackburn orientation relationship, i.e. (0002)Zr//(111)hydride and (1120)Zr//(1-10)hydride. The hydrides were also found to contain sigma 3 boundaries as well as local misorientations. (author)

Laser-Based Additive Manufacturing of Zirconium

Directory of Open Access Journals (Sweden)

Himanshu Sahasrabudhe

2018-03-01

Full Text Available Additive manufacturing of zirconium is attempted using commercial Laser Engineered Net Shaping (LENSTM technique. A LENSTM-based approach towards processing coatings and bulk parts of zirconium, a reactive metal, aims to minimize the inconvenience of traditional metallurgical practices of handling and processing zirconium-based parts that are particularly suited to small volumes and one-of-a-kind parts. This is a single-step manufacturing approach for obtaining near net shape fabrication of components. In the current research, Zr metal powder was processed in the form of coating on Ti6Al4V alloy substrate. Scanning electron microscopy (SEM and energy dispersive spectroscopy (EDS as well as phase analysis via X-ray diffraction (XRD were studied on these coatings. In addition to coatings, bulk parts were also fabricated using LENS™ from Zr metal powders, and measured part accuracy.

Quantitative analysis of nickel in zirconium and zircaloy

International Nuclear Information System (INIS)

Rastoix, M.

1957-01-01

A rapid spectrophotometric has been developed for determination of 10 to 1000 ppm of Ni in zirconium and zircaloy using dimethylglyoxime. Iron, copper, tin and chromium, do not interfere at the concentration usually present in zirconium and its alloys. (author) [fr

Melt propagation in dry core debris beds

International Nuclear Information System (INIS)

Dosanjh, S.S.

1989-01-01

During severe light water reactor accidents like Three Mile Island Unit 2, the fuel rods can fragment and thus convert the reactor core into a large particle bed. The postdryout meltdown of such debris beds is examined. A two-dimensional model that considers the presence of oxidic (UO 2 and ZrO 2 ) as well as metallic (e.g., zirconium) constituents is developed. Key results are that a dense metallic crust is created near the bottom of the bed as molten materials flow downward and freeze; liquid accumulates above the blockage and, if zirconium is present, the pool grows rapidly as molten zirconium dissolved both UO 2 and ZrO 2 particles; if the melt wets the solid, a fraction of the melt flows radially outward under the action of capillary forces and freezes near the radial boundary; in a nonwetting system, all of the melt flows into the bottom of the bed; and when zirconium and iron are in intimate contact and the zirconium metal atomic fraction is > 0.33, these metals can liquefy and flow out of the bed very early in the meltdown sequence

Immobilization of transition metal ions on zirconium phosphate monolayers

International Nuclear Information System (INIS)

Melezhik, A.V.; Brej, V.V.

1998-01-01

It is shown that ions of transition metals (copper, iron, vanadyl, titanium) are adsorbed on zirconium phosphate monolayers. The zirconium phosphate threshold capacity corresponds to substitution of all protons of hydroxyphosphate groups by equivalent amounts of copper, iron or vanadyl. Adsorption of polynuclear ions is possible in case of titanium. The layered substance with specific surface up to 300 m 2 /g, wherein ultradispersed titanium dioxide particles are intercalirated between zirconium-phosphate layers, is synthesized

Separated type nuclear superheating reactor

International Nuclear Information System (INIS)

Hida, Kazuki.

1993-01-01

In a separated type nuclear superheating reactor, fuel assemblies used in a reactor core comprise fuel rods made of nuclear fuel materials and moderator rods made of solid moderating materials such as hydrogenated zirconium. Since the moderating rods are fixed or made detachable, high energy neutrons generated from the fuel rods are moderated by the moderating rods to promote fission reaction of the fuel rods. Saturated steams supplied from the BWR type reactor by the fission energy are converted to high temperature superheated steams while passing through a steam channel disposed between the fuel rods and the moderating rods and supplied to a turbine. Since water is not used but solid moderating materials sealed in a cladding tube are used as moderation materials, isolation between superheated steams and water as moderators is not necessary. Further, since leakage of heat is reduced to improve a heat efficiency, the structure of the reactor core is simplified and fuel exchange is facilitated. (N.H.)

Viscoplastic behavior of zirconium alloys in the temperatures range 20 deg C - 400 deg C: characterization and modeling of strain ageing phenomena

International Nuclear Information System (INIS)

Graff, St.

2006-10-01

The anomalous strain rate sensitivity of zirconium alloys over the temperatures range 20-600 C has been widely reported in the literature. This unconventional behavior is related to the existence of strain ageing phenomenon which results from the combined action of thermally activated diffusion of foreign atoms to and along dislocation cores and the long range of dislocations interactions. The important role of interstitial and substitutional atoms in zirconium alloys, responsible for strain ageing and the lack of information about the domain where strain ageing is active have not been yet adequately characterized because of the multiplicity of alloying elements and chemical impurities. The aim of this work is to characterize experimentally the range of temperatures and strain rates where strain ageing is active on the macroscopic and mesoscopic scales. We propose also a predictive approach of the strain ageing effects, using the macroscopic strain ageing model suggested by McCormick (McCormick, 1988; Zhang et al., 2000). Specific zirconium alloys were elaborated starting from a crystal bar of zirconium with 2.2 wt% hafnium and very low oxygen content (80 wt ppm), called ZrHf. Another substitutional atom was added to the solid solution under the form of 1 wt% niobium. Some zirconium alloys were doped with oxygen, others were not. All of them were characterized by various mechanical tests (standard tensile tests, tensile tests with strain rate changes, relaxation tests with unloading). The experimental results were compared with those for the standard oxygen doped zirconium alloy (1300 wt ppm) studied by Pujol (Pujol, 1994) and called Zr702. The following experimental evidences of the age-hardening phenomena were collected and then modeled: 1) low and/or negative strain rate sensitivity around 200-300 C, 2) creep arrest at 200 C, 3) relaxation arrest at 200 C and 300 C, 4) plastic strain heterogeneities observed in laser extensometry on the millimeter scale

Viscoplastic behavior of zirconium alloys in the temperatures range 20 deg C - 400 deg C: characterization and modeling of strain ageing phenomena

Energy Technology Data Exchange (ETDEWEB)

Graff, St

2006-10-15

The anomalous strain rate sensitivity of zirconium alloys over the temperatures range 20-600 C has been widely reported in the literature. This unconventional behavior is related to the existence of strain ageing phenomenon which results from the combined action of thermally activated diffusion of foreign atoms to and along dislocation cores and the long range of dislocations interactions. The important role of interstitial and substitutional atoms in zirconium alloys, responsible for strain ageing and the lack of information about the domain where strain ageing is active have not been yet adequately characterized because of the multiplicity of alloying elements and chemical impurities. The aim of this work is to characterize experimentally the range of temperatures and strain rates where strain ageing is active on the macroscopic and mesoscopic scales. We propose also a predictive approach of the strain ageing effects, using the macroscopic strain ageing model suggested by McCormick (McCormick, 1988; Zhang et al., 2000). Specific zirconium alloys were elaborated starting from a crystal bar of zirconium with 2.2 wt% hafnium and very low oxygen content (80 wt ppm), called ZrHf. Another substitutional atom was added to the solid solution under the form of 1 wt% niobium. Some zirconium alloys were doped with oxygen, others were not. All of them were characterized by various mechanical tests (standard tensile tests, tensile tests with strain rate changes, relaxation tests with unloading). The experimental results were compared with those for the standard oxygen doped zirconium alloy (1300 wt ppm) studied by Pujol (Pujol, 1994) and called Zr702. The following experimental evidences of the age-hardening phenomena were collected and then modeled: 1) low and/or negative strain rate sensitivity around 200-300 C, 2) creep arrest at 200 C, 3) relaxation arrest at 200 C and 300 C, 4) plastic strain heterogeneities observed in laser extensometry on the millimeter scale

Determination of trace amounts of cadmium in zirconium and its alloys by graphite furnace AAS

International Nuclear Information System (INIS)

Takashima, Kyoichiro; Toida, Yukio

1994-01-01

Trace amount of cadmium in zirconium and its alloys was determined by graphite furnace atomic absorption spectrometry (GF-AAS) after ion exchange separation. A 2g chip sample was decomposed with 20ml of hydrofluoric acid (1+9) and a few drops of nitric acid. A trace amount of cadmium was separated from zirconium by strongly acidic cation-exchange resin (MCI GEL CK 08P) using 50ml of hydrochloric acid as an eluent. The solution was gently evaporated to dryness on an electric hot plate heater and under an infrared lamp. The residue was dissolved in 1ml of nitric acid (1+14) and diluted to 10ml in a volumetric glass flask with distilled water. Ten microliters of this solution was injected into a graphite furnace and then atomized at 2200degC for 4s in argon at a flow rate of 3.0l/min. Acids used in the analytical procedure were purified by azeotropic distillation and cation-exchange resin. The limit of determination (3σ BK ) for cadmium was 0.5ngCd/g and the relative standard deviation (RSD) at 1ngCd/g level was less than 20% for the GF-AAS. The accuracy of this technique was confirmed by NIST SRM 1643b (trace elements in water). (author)

Neutronic calculations of AFPR-100 reactor based on Spherical Cermet Fuel particles

International Nuclear Information System (INIS)

Benchrif, A.; Chetaine, A.; Amsil, H.

2013-01-01

Highlights: • AFPR-100 reactor considered as a small nuclear reactor without on-site refueling originally based on TRISO micro-fuel element. • The AFPR-100 reactor was re-designed using the new Spherical Cermet fuel element. • The adoption of the Cermet fuel instead of TRISO fuel reduces the core lifetime operation by 3.1 equivalent full power years. • We discussed the new micro-fuel element candidate for small and medium sized reactors. - Abstract: The Atoms For Peace Reactor (AFPR-100), as a 100 MW(e) without the need of on-site refueling, was originally based on UO2 TRISO fuel coated particles embedded in a carbon matrix directly cooled by light water. AFPR-100 is considered as a small nuclear reactor without open-vessel refueling which is proposed by Pacific Northwest National Laboratory (PNNL). An account of significant irradiation swelling in the silicon carbide fission product barrier coating layer of TRISO fuel element, a Spherical Cermet Fuel element has been proposed. Indeed, the new fuel concept, which was developed by PNNL, consists of changing the pyro-carbon and ceramic coatings that are incompatible with low temperature by Zirconium. The latter was chosen to avoid any potential Wigner energy effect issues in the TRISO fuel element. Actually, the purpose of this study is to assess the goal of AFPR-100 concept using the Cermet fuel; undeniably, the fuel core lifetime prediction may be extended for reasonably long period without on-site refueling. In fact, we investigated some neutronic parameters of reactor core by the calculation code SRAC95. The results suggest that the core fuel lifetime beyond 12 equivalent full power years (EFPYs) is possible. Hence, the adoption of Cermet fuel concept shows a core lifetime decrease of about 3.1 EFPY

Adaptation of fuel code for light water reactor with austenitic steel rod cladding

International Nuclear Information System (INIS)

Gomes, Daniel de Souza; Silva, Antonio Teixeira; Giovedi, Claudia

2015-01-01

Light water reactors were used with steel as nuclear fuel cladding from 1960 to 1980. The high performance proved that the use of low-carbon alloys could substitute the current zirconium alloys. Stainless steel is an alternative that can be used as cladding. The zirconium alloys replaced the steel. However, significant experiences in-pile occurred, in commercial units such as Haddam Neck, Indian Point, and Yankee experiences. Stainless Steel Types 347 and 348 can be used as cladding. An advantage of using Stainless Steel was evident in Fukushima when a large number of hydrogens was produced at high temperatures. The steel cladding does not eliminate the problem of accumulating free hydrogen, which can lead to a risk of explosion. In a boiling water reactor, environments easily exist for the attack of intergranular corrosion. The Stainless Steel alloys, Types 321, 347, and 348, are stabilized against attack by the addition of titanium, niobium, or tantalum. The steel Type 348 is composed of niobium, tantalum, and cobalt. Titanium preserves type 321, and niobium additions stabilize type 347. In recent years, research has increased on studying the effects of irradiation by fast neutrons. The impact of radiation includes changes in flow rate limits, deformation, and ductility. The irradiation can convert crystalline lattices into an amorphous structure. New proposals are emerging that suggest using a silicon carbide-based fuel rod cladding or iron-chromium-aluminum alloys. These materials can substitute the classic zirconium alloys. Once the steel Type 348 was chosen, the thermal and mechanical properties were coded in a library of functions. The fuel performance codes contain all features. A comparative analysis of the steel and zirconium alloys was made. The results demonstrate that the austenitic steel alloys are the viable candidates for substituting the zirconium alloys. (author)

Adaptation of fuel code for light water reactor with austenitic steel rod cladding

Energy Technology Data Exchange (ETDEWEB)

Gomes, Daniel de Souza; Silva, Antonio Teixeira, E-mail: dsgomes@ipen.br, E-mail: teixeira@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (POLI/USP), Sao Paulo, SP (Brazil). Lab. de Analise, Avaliacao e Gerenciamento de Risco

2015-07-01

Light water reactors were used with steel as nuclear fuel cladding from 1960 to 1980. The high performance proved that the use of low-carbon alloys could substitute the current zirconium alloys. Stainless steel is an alternative that can be used as cladding. The zirconium alloys replaced the steel. However, significant experiences in-pile occurred, in commercial units such as Haddam Neck, Indian Point, and Yankee experiences. Stainless Steel Types 347 and 348 can be used as cladding. An advantage of using Stainless Steel was evident in Fukushima when a large number of hydrogens was produced at high temperatures. The steel cladding does not eliminate the problem of accumulating free hydrogen, which can lead to a risk of explosion. In a boiling water reactor, environments easily exist for the attack of intergranular corrosion. The Stainless Steel alloys, Types 321, 347, and 348, are stabilized against attack by the addition of titanium, niobium, or tantalum. The steel Type 348 is composed of niobium, tantalum, and cobalt. Titanium preserves type 321, and niobium additions stabilize type 347. In recent years, research has increased on studying the effects of irradiation by fast neutrons. The impact of radiation includes changes in flow rate limits, deformation, and ductility. The irradiation can convert crystalline lattices into an amorphous structure. New proposals are emerging that suggest using a silicon carbide-based fuel rod cladding or iron-chromium-aluminum alloys. These materials can substitute the classic zirconium alloys. Once the steel Type 348 was chosen, the thermal and mechanical properties were coded in a library of functions. The fuel performance codes contain all features. A comparative analysis of the steel and zirconium alloys was made. The results demonstrate that the austenitic steel alloys are the viable candidates for substituting the zirconium alloys. (author)

The chemical vapor deposition of zirconium carbide onto ceramic substrates

International Nuclear Information System (INIS)

Glass A, John Jr.; Palmisiano, Nick Jr.; Welsh R, Edward

1999-01-01

Zirconium carbide is an attractive ceramic material due to its unique properties such as high melting point, good thermal conductivity, and chemical resistance. The controlled preparation of zirconium carbide films of superstoichiometric, stoichiometric, and substoichiometric compositions has been achieved utilizing zirconium tetrachloride and methane precursor gases in an atmospheric pressure high temperature chemical vapor deposition system

Fuel gases generation in the primary contention during a coolant loss accident in a nuclear power plant with reactor type BWR

International Nuclear Information System (INIS)

Salaices, M.; Salaices, E.; Ovando, R.; Esquivias, J.

2011-11-01

During an accident design base of coolant loos, the hydrogen gas can accumulate inside the primary contention as a result of several generation mechanisms among those that are: 1) the reaction metal-water involving the zirconium of the fuel cladding and the reactor coolant, 2) the metals corrosion for the solutions used in the emergency cooling and dew of the contention, and 3) the radio-decomposition of the cooling solutions of post-accident emergency. In this work the contribution of each generation mechanism to the hydrogen total in the primary contention is analyzed, considering typical inventories of zirconium, zinc, aluminum and fission products in balance cycle of a reactor type BWR. In the analysis the distribution model of fission products and hydrogen production proposed in the regulator guide 1.7, Rev. 2 of the US NRC was used. The results indicate that the mechanism that more contributes to the hydrogen generation at the end of a period of 24 hours of initiate the accident is the radio-decomposition of the cooling solutions of post-accident emergency continued by the reaction metal-water involving the zirconium of the fuel cladding with the reactor coolant, and lastly the aluminum and zinc oxidation present in the primary contention. However, the reaction metal-water involving the zirconium of the fuel cladding and the reactor coolant is the mechanism that more contributes to the hydrogen generation in the first moments after the accident. This study constitutes the first part of the general analysis of the generation, transport and control of fuel gases in the primary contention during a coolant loss accident in BWRs. (Author)

Preparation and certification of certified reference materials JAERI-Z21, Z22 and Z23 for analysis of zirconium and its alloys

International Nuclear Information System (INIS)

Takashima, Kyoichiro

1991-03-01

The Sub-Committee on Chemical Analysis of Nuclear Materials was organized in April 1987, under the Committee on Analytical Chemistry of Nuclear Fuels and Reactor Materials, JAERI, for renewal of certified reference materials of zirconium base alloys and zirconium metal. Collaborative analysis was carried out among ten participating laboratories for the certification of the JAERI CRMs Z21 to Z23. As a results of the collaborative works, the certified values for sixteen elements (Sn, Fe, Ni, Cr, Hf, Al, Si, Co, Cu, Ti, Mn, Pb, U, Cd, B and W) in the CRMs were given. In this report, preparation of raw materials, homogeneity test, chemical analysis for certification by collaborative works during April 1987 to March 1990 are described. (author)

Development of rolled joints for zirconium-2.5 wt % niobium pressure tubes

International Nuclear Information System (INIS)

Madhusoodanan, K.; Sinha, R.K.; Samuel, K.A.; Joeman, V.

1992-01-01

Due to its higher strength and lower deuterium pick-up rate, as compared to the existing cold worked zircaloy-2 material, cold worked zirconium-2.5 wt% niobium (Zr-2.5%Nb) alloy is to be used as the pressure tube material in all forthcoming Indian PHWRs starting with KAPP-2. These pressure tubes, which carry the fuel bundles are to be joined to the S.S 403 end-fittings through rolled joints. Since the new pressure tubes have a lower wall thickness and higher room temperature yield stress, than zircaloy-2 tubes the design parameters of the rolled joint had to be developed afresh. Further, since Zr-2.5%Nb is susceptible to delayed hydride cracking, it is necessary to limit the residual stress near the rolled joint to a minimum. Since the high residual stress is due to the initial assembly clearance between the pressure tube and end-fitting, a modified rolled joint had to be developed, referred to as zero clearance rolled joint. This paper provides details of the work carried out at Reactor Engineering Division of Bhabha Atomic Research Centre, Bombay towards the development of the design of the rolled joint as well as the tooling and procedures required for achieving zero-clearance fit-ups at site. The requirements to be met by the Zr-2.5% Nb pressure tubes for achieving acceptable rolled joints are highlighted. (author). 5 refs., 6 figs., 3 tabs

Young's modulus of crystal bar zirconium and zirconium alloys (zircaloy-2, zircaloy-4, zirconium-2.5wt% niobium) to 1000 K

International Nuclear Information System (INIS)

Rosinger, H.E.; Ritchie, I.G.; Shillinglaw, A.J.

1975-09-01

This report contains experimentally determined data on the dynamic elastic moduli of zircaloy-2, zircaloy-4, zirconium-2.5wt% niobium and Marz grade crystal bar zirconium. Data on both the dynamic Young's moduli and shear moduli of the alloys have been measured at room temperature and Young's modulus as a function of temperature has been determined over the temperature range 300 K to 1000 K. In every case, Young's modulus decreases linearly with increasing temperature and is expressed by an empirical equation fitted to the data. Differences in Young's modulus values determined from specimens with longitudinal axes parallel and perpendicular to the rolling direction are small, as are the differences between Young's moduli determined from strip, bar stock and fuel sheathing. (author)

In vitro assessment of artifacts induced by titanium, titanium-zirconium and zirconium dioxide implants in cone-beam computed tomography.

Science.gov (United States)

Sancho-Puchades, Manuel; Hämmerle, Christoph H F; Benic, Goran I

2015-10-01

The aim of this study was to test whether or not the intensity of artifacts around implants in cone-beam computed tomography (CBCT) differs between titanium, titanium-zirconium and zirconium dioxide implants. Twenty models of a human mandible, each containing one implant in the single-tooth gap position 45, were cast in dental stone. Five test models were produced for each of the following implant types: titanium 4.1 mm diameter (Ti4.1 ), titanium 3.3 mm diameter (Ti3.3 ), titanium-zirconium 3.3 mm diameter (TiZr3.3 ) and zirconium dioxide 3.5-4.5 mm diameter (ZrO3.5-4.5 ) implants. For control purposes, three models without implants were produced. Each model was scanned using a CBCT device. Gray values (GV) were recorded at eight circumferential positions around the implants at 0.5 mm, 1 mm and 2 mm from the implant surface (GVT est ). GV were assessed in the corresponding volumes of interest (VOI) in the control models without implants (GVC ontrol ). Differences of gray values (ΔGV) between GVT est and GVC ontrol were calculated as percentages. One-way ANOVA and post hoc tests were applied to detect differences between implant types. Mean ΔGV for ZrO3.5-4.5 presented the highest absolute values, generally followed by TiZr3.3 , Ti4.1 and Ti3.3 implants. The differences of ΔGV between ZrO3.5-4.5 and the remaining groups were statistically significant in the majority of the VOI (P ≤ 0.0167). ΔGV for TiZr3.3 , Ti4.1 and Ti3.3 implants did not differ significantly in the most VOI. For all implant types, ΔGV showed positive values buccally, mesio-buccally, lingually and disto-lingually, whereas negative values were detected mesially and distally. Zirconium dioxide implants generate significantly more artifacts as compared to titanium and titanium-zirconium implants. The intensity of artifacts around zirconium dioxide implants exhibited in average the threefold in comparison with titanium implants. © 2014 John Wiley & Sons A/S. Published by John Wiley

Spectrophotometric titration of sulfates in the presence of zirconium

International Nuclear Information System (INIS)

Kuznetsov, V.V.; Kotova, S.S.; Molokanova, L.G.; Chekmarev, A.M.; Yagodin, G.A.

1978-01-01

The procedure has been proposed for express determination of sulphate ions in the presence of zirconium by spectrophotometric titration with the use of barium chloride and nitrochromazo as an indicator. The procedure is based on bonding zirconium into a more stable complex with EDTA (ethylenediaminotetraacetic acid). The presence of excess of EDTA and zirconium (4) complexonate in the solution being titrated does not affect the titration curve shape and the character of break on the curve in the equivalence point. A complete demasking of SO 4 2- is observed in the case of 1O-fold excess of EDTA with respect to zirconium (4). Statistic evaluation of the method has shown that the results of titration can be distorted by chance errors only

The atomic energy basic law

International Nuclear Information System (INIS)

1977-01-01

The law establishes clearly the principles that Japan makes R and D, and utilizations of atomic energy only for the peaceful purposes. All the other laws and regulations concerning atomic energy are based on the law. The first chapter lays down the above mentioned objective of the law, and gives definitions of basic concepts and terms, such as atomic energy, nuclear fuel material, nuclear source material, nuclear reactor and radiation. The second chapter provides for the establishment of Atomic Energy Commission which conducts plannings and investigations, and also makes decisions concerning R and D, and utilizations of atomic energy. The third chapter stipulates for establishment of two government organizations which perform R and D of atomic energy developments including experiments and demonstrations of new types of reactors, namely, Atomic Energy Research Institute and Power Reactor and Nuclear Fuel Development Corporation. Chapters from 4th through 8th provide for the regulations on development and acquisition of the minerals containing nuclear source materials, controls on nuclear fuel materials and nuclear reactors, administrations of the patents and inventions concerning atomic energy, and also prevention of injuries due to radiations. The last 9th chapter requires the government and its appointee to compensate the interested third party for damages in relation to the exploitation of nuclear source materials. (Matsushima, A.)

Study of the uranium-zirconium diffusion; Etude de la diffusion uranium-zirconium

Energy Technology Data Exchange (ETDEWEB)

Adda, Y; Mairy, C; Bouchet, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1957-07-01

The intermetallic diffusion of uranium fuel and zirconium used as cladding is studied. Intermetallic diffusion can occur during the cladding of uranium rods and uranium can penetrate the zirconium cladding. Different parameters are involved in this mechanism as structure and mechanical properties of the diffusion area as well as presence of impurities in the metal. The uses of different analysis techniques (micrography, Castaing electronic microprobe, microhardness and autoradiography) have permitted to determine with great accuracy the diffusion coefficient in gamma phase (body centered cubic system) and the results have given important information on the intermetallic diffusion mechanisms. The existence of the Kirkendall effect in the U-Zr diffusion is also an argument in favor of the generality of the diffusion mechanism by vacancies in body centered cubic system. (M.P.)

Zirconium ignition in exposed fuel channel

Energy Technology Data Exchange (ETDEWEB)

Elias, E., E-mail: merezra@technion.ac.il; Hasan, D.; Nekhamkin, Y.

2015-05-15

Highlights: • We demonstrate the idea of runaway zirconium–steam reactions in severe accidents in today's LWRs. • We predict the thermal-hydraulics conditions relevant to cladding oxidation in an exposed fuel channel of a partially uncovered core. • The Semenov theory of metal combustion is extended to define a criterion for runaway oxidation reaction in fuel cladding. - Abstract: A theoretical model based on simultaneous solution of the heat and mass transfer equations is developed for predicting the rate of thermo-chemical reaction between zirconium cladding and a hot steam environment. Ignition conditions relevant to cladding oxidation in an exposed fuel channel of a partially uncovered core are predicted based on the theory of metal combustion. A range of decay power, convective heat transfer coefficients, and initial temperatures leading to uncontrolled runaway cladding oxidation is identified. The model could be readily integrated as part of a fuel channel analysis code for predicting possible outcomes of different accident mitigation procedures in light water nuclear reactors under LOCA conditions.

Amendment of Atomic Energy Basic Law and the development of Atomic Energy Administration

International Nuclear Information System (INIS)

Ochi, Kenji

1978-01-01

This article explains the key points of the major development of Atomic Energy Administration recently made by amendments of Atomic Energy Basic Law and other two relating laws. These amendments passed through the Diet and were enacted on 7th, June, 1978. The aim of them is focussed on reinforcement and rearrangement of safety controls on nuclear reactors. Previously, although the approval of the installation plan with basic designs of a nuclear reactor has been done by Prime Minister, further approvals of detailed designs and process of construction works, as well as inspections before and after operation have been conducted by each responsible minister, respectively. That is, those controls for power reactors have been within jurisdiction of minister of Trade and Industry, and for nuclear ships' reactors minister of Transportation has been responsible. Under the new system, above mentioned ministers continue to exercise almost same controls over reactors within their jurisdiction respectively, however the new laws have established so-called ''double check'' principle in that: when each responsible minister approves the installation, detailed designs and further stages of construction and operation of the reactor, he should hear and pay a great regard for opinions of Atomic Energy Commission and Atomic Energy Safety Commission. The latter is newly established organization which has similar status and authority to the former. (J.P.N.)

Uranium (Vi) sorption onto zirconium diphosphate chemically modified

International Nuclear Information System (INIS)

Garcia G, N.; Ordonez R, E.

2010-10-01

This work deals with the uranium (Vi) speciation after sorption onto zirconium diphosphate (ZrP 2 O 7 ) surface, hydrated and in a surface modified with organic acids. Oxalic and citric acids were chosen to modify the ZrP 2 O 7 surface because they have poly carboxylic groups and they mimic the organic matter in nature. Thus the interest of this work is to evaluate the uranium (Vi) sorption edge at different s ph values in natural and modified surfaces. The luminescence technique (fluorescence and phosphorescence, respectively) was used for the quantification and speciation of uranyl sorbed at the zirconium diphosphate interface. The fluorescence experiment, showed that adsorption of uranyl on surface of zirconium diphosphate tends to 100%. The speciation shows that there are different complexes in surface which were formed between zirconium diphosphate and uranyl, since it is produced a displacement of wavelength in fluorescence spectra of each system. (Author)

A study of a production process for hafnium-free zirconium from zircon

International Nuclear Information System (INIS)

Ratanalert, N.

1985-01-01

The purpose of this experiment was to extract and purify the zirconium from zircon. The effects of time of extraction and stripping of zirconium, concentration of feed solution, concentration of hydrochloric acid in stripping process, equilibrium curve of extraction of zirconium and hafnium and equilibrium curve of stripping zirconium or scrubbing hafnium were studied from standard zirconium and hafnium. The results, subsequently were applied to the extraction procedures for zirconium from zircon. Minus 100 mesh zircon was fused with sodium hydroxide in the ratio of 1 : 6 at 700 degree C for l hour. After fusion the zirconate was leached with water and dissolved in hot concentrated hydrochloric acid. Zirconyl chloride octahydrate crystallized out when the solution was cooled. An agueons solution of zirconyl chloride was used as the feed to the hexone - thiocyanate solvent extraction process. This was prepared by dissolving zirconyl chloride octahydrate crystal in waster. This zirconium feed solution in 1 M HCl and 1 M N H 4 CNS was extracted with 2.7 m N H 4 CNS in hexone and then stripped with 3.6 M HCl the aqueous phase was got rid of thiocyanate ion by extracting with pure hexone, then the zirconium in aqueous phase was precipitated with sulfuric acid and ammonium hydroxide at pH 1.8 - 2.0 and zirconium oxide was obtained by ignition at 700 degree C. The process could be modified to improve the purity of zirconium by using cation exchange resin to get rid of thiocyanate ion after solvent extraction process

Cobalt-60 production in the BN-350 fast power reactor

International Nuclear Information System (INIS)

Zvonarev, A.V.; Korobejnikov, V.V.; Matveenko, I.P.

1994-01-01

A possibility of Co-60 isotope production in the BN-350 fast reactor was considered. A special irradiating device, which is an assembly with a central hole, where a container containing cobalt and zirconium hydride is placed. The irradiating device tested permits generating 60 Co with specific activity of 100 Ci/g

Contribution to the study of the physico-chemical mechanisms of metallic cation extraction by alkylphosphoric acids. Extraction of zirconium (IV) by di-2-ethylhexyl phosphoric acid (DEHPA)

International Nuclear Information System (INIS)

Carbonnier, J.-L.

1979-02-01

Extraction of zirconium, especially at high concentration (0.1M), by dodecane diluted DEHPA (HA) from hydrochloric or nitric aqueous phases of 0.1 to 10 M acidity was studied. The composition, structure and polymerisation of the complexes extracted were determined by chemical analysis, viscosimetry, infrared spectrometry and light scattering. A Zr(OH) 2 A 2 .2HNO 3 , type structure is proposed for these complexes instead of the generally accepted form: Zr(OH) 2 (NO 3 ) 2 .2HA. Similarly in hydrochloric solution: Zr(OH) 2 A 2 .2HCl. Polymerisation in the organic phase results from the juxtaposition of two factors; firstly zirconium saturation (formation of bridges by DEHPA between zirconium atoms) and secondly the nature the equeous phase. In slightly acid hydrochloric solution (pH = 1.3) the aqueous plymers of zirconium are extracted in the organic phase as polynuclear complexes; in nitric solution no polynuclear complexes are observed but the nitric acid molecules extracted set up hydrogen bonds which explain the increased viscosity and gelification of the organic phases [fr

Translucency and Strength of High Translucency Monolithic Zirconium Oxide Materials

Science.gov (United States)

2016-05-17

Zirconium -Oxide Materials presented at/published to the Journal of General Dentistry with MDWI 41-108, and has been assigned local file #16208. 2...Zirconia-Oxide Materials 6. TITLE OF MATERIAL TO BE PUBLISHED OR PRESENTED: Translucency and Strength of High-Translucency Monolithic Zirconium -Oxide...OBSOLETE 48. DATE Page 3 of 3 Pages Translucency and Strength of High-Translucency Monolithic Zirconium -Oxide Materials Abstract Dental materials

Design and safety considerations for the 10 MW(t) multipurpose TRIGA reactor in Thailand

International Nuclear Information System (INIS)

Razvi, J.; Bolin, J.M.; Saurwein, J.J.; Whittemore, W.L.; Proongmuang, S.

1999-01-01

General Atomics (GA) is constructing the Ongkharak Nuclear Research Center (ONRC) near Bangkok, Thailand for the Office of Atomic Energy for Peace. The ONRC complex includes the following: A multipurpose 10 MW(t) research reactor; An Isotope Production Facility; Centralized Radioactive Waste Processing and Storage Facilities. The Center is being built 60-km northeast of Bangkok, with a 10 MW(t) TRIGA type research reactor as the centerpiece. Facilities are included for neutron transmutation doping of silicon, neutron capture therapy neutron beam research and for production of a variety of radioisotopes. The facility will also be utilized for applied research and technology development as well as training in reactor operations, conduct of experiments and in reactor physics. The multipurpose, pool-type reactor will be fueled with high-density (45 wt%), low-enriched (19.7 wt%) uranium-erbium-zirconium-hydride (UErZrH) fuel rods, cooled and moderated by light water, and reflected by beryllium and heavy water. The general arrangement of the reactor and auxiliary pool structure allows irradiated targets to be transferred entirely under water from their irradiation locations to the hot cell, then pneumatically transferred to the adjacent Isotope Production Facility for processing. The core configuration includes 4 x 4 array standard TRIGA fuel clusters, modified clusters to serve as fast-neutron irradiation facilities, control rods and an in-core Ir-192 production facility. The active core is reflected on two sides by beryllium and on the other two sides by D 2 O. Additional irradiation facilities are also located in the beryllium reflector blocks and the D 2 O reflector blanket. The fuel provides the fundamental safety feature of the ONRC reactor, and as a result of all the well established accident-mitigating characteristics of the UErZrH fuel itself (large prompt negative temperature coefficient of reactivity, fission product retention and chemical stability), a

PROCESS FOR DISSOLVING BINARY URANIUM-ZIRCONIUM OR ZIRCONIUM-BASE ALLOYS

Science.gov (United States)

Jonke, A.A.; Barghusen, J.J.; Levitz, N.M.

1962-08-14

A process of dissolving uranium-- zirconium and zircaloy alloys, e.g. jackets of fuel elements, with an anhydrous hydrogen fluoride containing from 10 to 32% by weight of hydrogen chloride at between 400 and 450 deg C., preferably while in contact with a fluidized inert powder, such as calcium fluoride is described. (AEC)

High-resolution characterization of oxidation mechanism of zirconium nuclear fuel cladding alloys

International Nuclear Information System (INIS)

Hu, J.; Lozano-Perez, S.; Grovenor, C.

2015-01-01

Full text of publication follows. Zirconium alloys are used extensively as cladding materials in modern light water reactors to separate the uranium dioxide (UO 2 ) fuel rods and the coolant water in order to prevent the escape of radioactive fission products whilst maintaining heat transfer to the coolant. With increasing demand for high burn-up in modern nuclear reactors, environmental degradation of these alloys is now the life limiting factor for fuel assemblies. As part of the MUZIC-2 collaboration studying oxidation and hydrogen pickup in Zr alloys, several high resolution analysis techniques have been used to study the microstructure of a range of commercial and developmental Zr alloys. The sample used for this investigation was prepared from a Westinghouse TM developmental alloy with composition of Zr-0.9Nb-0.01Sn-0.08Fe (wt %) in the recrystallized condition. The sample was oxidised in an autoclave at EDF Energy under simulated PWR water conditions at 360 C. degrees for 360 days. Using Transmission Electron Microscope (TEM), we have studied the development of the equiaxed-columnar-equiaxed grain structure, and observe that the columnar grains are both longer and show a stronger preferred texture in more corrosion-resistant alloys. Fresnel imaging revealed the existence of both parallel interconnected pores and some vertically interconnected pores along the columnar oxide grain boundaries, which become more disconnected near the metal-oxide interface. Electron Energy Loss Spectroscopy (EELS) provided accurate quantitative analysis of the oxygen concentration across the interface, identifying the existence of local regions of stoichiometric ZrO and Zr 3 O 2 with varying thickness. These observations will be discussed in the context of current models for oxidation in zirconium alloys. (authors)

The determination of some impurities in zirconium metal by instrumental neutron activation analysis

International Nuclear Information System (INIS)

Eddy, B.T.; Pearton, D.C.G.; Watterson, J.J.

1976-01-01

This report describes the work done on the development of an instrumental neutron-activation method for the analysis of impurities in reactor-grade zirconium. Nine samples were analysed, and the results were compared with those obtained by other techniques. No statistically significant differences were observed for ten of the twelve elements that could possibly be determined by instrumental neutron-activation analysis. Cadmium cannot be determined at the 0,5 p.p.m. level, and there is doubt about the comparative values recorded for aluminium. The precision of the measurement by direct instrumental neutron-activation analysis ranges from 1,4 per cent for tungsten to 17 per cent for chromium [af

Capture of Tritium Released from Cladding in the Zirconium Recycle Process

Energy Technology Data Exchange (ETDEWEB)

Spencer, Barry B [ORNL; Bruffey, Stephanie H [ORNL; DelCul, Guillermo Daniel [ORNL; Walker, Trenton Baird [ORNL

2016-08-31

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.

Capture of Tritium Released from Cladding in the Zirconium Recycle Process

Energy Technology Data Exchange (ETDEWEB)

Bruffey, Stephanie H [ORNL; Spencer, Barry B [ORNL; DelCul, Guillermo Daniel [ORNL

2016-08-31

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.

Methods for determination of zirconium in titanium alloys

International Nuclear Information System (INIS)

1985-01-01

Two methods for determining zirconium content in titanium alloys are specified in this standard. One is the ion-exchange/mandelic acid gravimetry for Zr content below 20 % down to 1 % while the other is the mandelic acid gravimetry for Zr content below 20 % down to 0.5 %. In the former, a specimen is decomposed by hydrochloric acid and hydrofluoric acid. After substances such as titanium are oxidized by adding nitric acid, the liquid is adjusted into a 4N hydrochloric acid - gN hydrofluoric acid solution, which is them passed through an ion-exchange column. The niobium and tantalum contents are absorbed while the titanium and zirconium contents flow out. Perchloric acid and sulfuric acid are poured in the solution to remove hydrofluoric acid. Aqueous ammonia is added to produce hydroxide of titanium and zirconium, which is then filtered out. The hydroxyde is dissolved in hydrochloric acid, and mandelic acid is poured to precipitate the zirconium content. The precipitate is ignited and the weight of the oxide formed is measured. The coprecipitated titanium content is determined by the absorptiometric method using hydrogen peroxide. Finally, the weight of the oxide is corrected. In the latter determination method, on the other hand, only several steps of the above procedure are used, namely, decomposition by hydrochloric acid, precipitation of zirconium, ignition of precipitate, measurement of oxide weight and weight correction. (Nogami, K.)

Corrosion resistant zirconium alloys prepared by powder metallurgy

International Nuclear Information System (INIS)

Wojeik, C.C.

1984-01-01

Pure zirconium and zirconium 2.5% niobium were prepared by powder metallurgy. The powders were prepared directly from sponge and consolidated by cold isostatic pressing and sintering. Hot isostatic pressing was also used to obtain full density after sintering. For pure zirconium the effects of particle size, compaction pressure, sintering temperature and purity were investigated. Fully densified zirconium and Zr-2.5%Nb exhibited tensile properties comparable to cast material at room temperature and 300 0 F (149 0 C). Pressed and sintered material having density of 94-99% had slightly lower tensile properties. Corrosion tests were performed in boiling 65% H/sub 2/SO/sub 4/, 70% HNO/sub 3/, 20% HCl and 20% HCl + 500 ppm FeCl/sub 3/ (a known pitting solution). For fully dense material the observed corrosion behavior was nearly equivalent to cast material. A slightly higher rate of attack was observed for samples which were only 94-99% dense. Welding tests were also performed on zirconium and Zr-2.5%Nb alloy. Unlike P/M titanium alloys, these materials had good weldability due to the lower content of volatile impurities in the powder. A slight amount of weld porosity was observed but joint efficiencies were always not 100%, even for 94-99% density samples. Several practical applications of the P/M processed material will be briefly described

Investigation on the corrosion resistance of zirconium in nitric acid

International Nuclear Information System (INIS)

Fauvet, P.; Mur, P.

1990-01-01

Zirconium in nitric solutions exhibits an excellent corrosion resistance in the passive state, and a mediocre corrosion resistance in the unpassive state with risk of stress corrosion cracking. Results of the influence of some parameters (medium, potential, temperature, stress, friction, metallurgical structure and surface state) on zirconium passivation are presented. Zirconium remains passive in a large range of HNO 3 concentration (at least up to 14.4N), in the presence of oxidizing ions (Cr 4 , Ce 4 ), in a spent fuel dissolution solution. Zirconium is depassived by friction at high speed and pressure, by platinum coupling in boiling 14.4N HNO 3 with or without stress, or by imposed deformation speed under high potential. (A.B.)

Effect of Burnable Absorbers on Inert Matrix Fuel Performance and Transuranic Burnup in a Low Power Density Light-Water Reactor

Directory of Open Access Journals (Sweden)

Geoff Recktenwald

2013-04-01

Full Text Available Zirconium dioxide has received particular attention as a fuel matrix because of its ability to form a solid solution with transuranic elements, natural radiation stability and desirable mechanical properties. However, zirconium dioxide has a lower coefficient of thermal conductivity than uranium dioxide and this presents an obstacle to the deployment of these fuels in commercial reactors. Here we show that axial doping of a zirconium dioxide based fuel with erbium reduces power peaking and fuel temperature. Full core simulations of a modified AP1000 core were done using MCNPX 2.7.0. The inert matrix fuel contained 15 w/o transuranics at its beginning of life and constituted 28% of the assemblies in the core. Axial doping reduced power peaking at startup by more than ~23% in the axial direction and reduced the peak to average power within the core from 1.80 to 1.44. The core was able to remain critical between refueling while running at a simulated 2000 MWth on an 18 month refueling cycle. The results show that the reactor would maintain negative core average reactivity and void coefficients during operation. This type of fuel cycle would reduce the overall production of transuranics in a pressurized water reactor by 86%.

Thermotransport of nitrogen and oxygen in β-zirconium

NARCIS (Netherlands)

Vogel, D.L.; Rieck, G.D.

1971-01-01

An investigation of thermotransport of nitrogen in ß-zirconium is reported. Using a method previously described, the heat of transport turned out to be 25.1 kcal/mole with a standard deviation of 2.5 kcal/mole. The formerly published value of the heat of transport of oxygen in ß-zirconium, viz. 20

Dielectric properties of zirconium dioxide-based ceramics

International Nuclear Information System (INIS)

Vladimirova, O.S.; Gruzdev, A.I.; Koposova, Z.L.; Lyutsareva, L.A.

1985-01-01

This paper studies the dielectric properties of materials based on stabilized zirconium dioxide with Co 3 O 4 additions possessing a high temperature-coefficient of resistance. These materials are promising for manufacturing resistance temperature gages that work under an oxidizing atmosphere at 370-1270 degrees K. The obtained results indicate the possibility of developing temperature gases possessing highsensitivity from stabilized zirconium dioxide with Co 3 O 4 additions

METHOD OF IMPROVING CORROSION RESISTANCE OF ZIRCONIUM

Science.gov (United States)

Shannon, D.W.

1961-03-28

An improved intermediate rinse for zirconium counteracts an anomalous deposit that often results in crevices and outof-the-way places when ordinary water is used to rinse away a strong fluoride etching solution designed to promote passivation of the metal. The intermediate rinse, which is used after the etching solution and before the water, is characterized by a complexing agent for fluoride ions such as aluminum or zirconium nitrates or chlorides.

Accelerated irradiation growth of zirconium alloys

International Nuclear Information System (INIS)

Griffiths, M.; Gilbert, R.W.; Fidleris, V.

1989-01-01

This paper discusses how sponge zirconium and Zr-2.5 wt% Nb, Zircaloy, or Excel alloys all exhibit accelerated irradiation growth compared with high-purity crystal-bar zirconium for irradiation temperatures between 550 to 710 K and fluences between 0.1 to 10 x 10 25 n · m -2 (E > 1 MeV). There is generally an incubation period or fluence before the onset of accelerated or breakaway growth, which is dependent on the particular material being irradiated, its metallurgical condition before irradiation, and the irradiation temperature. Transmission electron microscopy has shown that there is a correlation between accelerated irradiation growth and the appearance of c-component vacancy loops on basal planes. Measurements in some specimens indicate that a significant fraction of the strain can be directly attributed to the loops themselves. There is considerable evidence to show that their formation is dependent both on the specimen purity and on the irradiation temperature. Materials that have a high interstitial-solute content contain c-component loops and exhibit high growth rates even at low fluences ( 2 :5 n · m -2 , E > 1 MeV). For sponge zirconium and the Zircaloys, c-component loop formation and the associated acceleration of growth (breakaway) during irradiation occurs because the intrinsic interstitial solute (mainly, oxygen, carbon and nitrogen) in the zirconium matrix is supplemented by interstitial iron, chromium, and nickel from the radiation-induced dissolution of precipitates. (author)

Primary Stability of Zirconium vs Titanium Implants: An In Vitro Comparison

Science.gov (United States)

2015-06-05

of any copyrighted material in the thesis manuscript entitled: Primary Stability of Zirconium vs Titanium Implants: An In Vitro Comparison Is...Uniformed Services University Date: 02/20/2015 Primary Stability of Zirconium vs Titanium Implants: An In Vitro Comparison By...the thesis manuscript entitled: Primary Stability of Zirconium vs Titanium Implants: An In Vitro Comparison Is appropriately acknowledged

Separation of zirconium from hafnium by ion exchange

Energy Technology Data Exchange (ETDEWEB)

Felipe, Elaine C.B.; Palhares, Hugo G.; Ladeira, Ana Claudia Q., E-mail: elainecfelipe@yahoo.com.br, E-mail: hugopalhares@gmail.com, E-mail: ana.ladeira@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2013-07-01

Zirconium and hafnium are two of the most important metals for the nuclear industry. Hafnium occurs in all zirconium ores usually in the range 2 - 3%. However, for the most nuclear industry applications, it is necessary to use a zirconium of extremely pure level. The current work consists in the separation of zirconium and hafnium by the ion exchange method in order to obtain a zirconium concentrate of high purity. The zirconium and hafnium liquors were produced from the leaching of the Zr(OH){sub 4} and Hf(OH){sub 4} with nitric acid for 24 hours. From these two liquors it was prepared one solution containing 7.5 x 10{sup -2} mol L{sup -1} of Zr and 5.8 x 10{sup -3} mol L{sup -1} of Hf with acidity of 1 M. Ion exchange experiments were carried out in batch with the resins Dowex 50WX4, Dowex 50WX8 100, Dowex 50WX8 50, Amberlite IR-120 and Marathon C at constant temperature 28 deg C. Other variables such as, acidity and agitation were kept constant. The data were adjusted to Langmuir equation in order to calculate the maximum loading capacity (q{sub max}) of the resins, the distribution coefficient (K{sub d}) for Zr and Hf and the separation factor (α{sub Hf}{sup Zr} ). The results of maximum loading capacity (q{sub max}) for Zr and Hf, in mmol g{sup -}1, showed that the most suitable resins for columns experiments are: Dowex 50WX4 50 (q{sub max} Z{sub r} = 2.21, Hf = 0.18), Dowex 50WX8 50 (q{sub max} Zr = 1.89, Hf = 0.13) and Amberlite (q{sub max} Zr = 1.64, Hf = 0.12). However, separations factors, α{sub Hf}{sup Zr}, showed that the resins are not selective. (author)

Improving the neutronic characteristics of a boiling water reactor by using uranium zirconium hydride fuel instead of uranium dioxide fuel

Energy Technology Data Exchange (ETDEWEB)

Galahom, Ahmed Abdelghafar [Higher Technological Institute, Ramadan (Egypt)

2016-06-15

The present work discusses two different models of boiling water reactor (BWR) bundle to compare the neutronic characteristics of uranium dioxide (UO{sub 2}) and uranium zirconium hydride (UZrH{sub 1.6}) fuel. Each bundle consists of four assemblies. The BWR assembly fueled with UO{sub 2} contains 8 × 8 fuel rods while that fueled with UZrH{sub 1.6} contains 9 × 9 fuel rods. The Monte Carlo N-Particle Transport code, based on the Mont Carlo method, is used to design three dimensional models for BWR fuel bundles at typical operating temperatures and pressure conditions. These models are used to determine the multiplication factor, pin-by-pin power distribution, axial power distribution, thermal neutron flux distribution, and axial thermal neutron flux. The moderator and coolant (water) are permitted to boil within the BWR core forming steam bubbles, so it is important to calculate the reactivity effect of voiding at different values. It is found that the hydride fuel bundle design can be simplified by eliminating water rods and replacing the control blade with control rods. UZrH{sub 1.6} fuel improves the performance of the BWR in different ways such as increasing the energy extracted per fuel assembly, reducing the uranium ore, and reducing the plutonium accumulated in the BWR through burnup.

Particle bed reactor nuclear rocket concept

International Nuclear Information System (INIS)

Ludewig, H.

1991-01-01

The particle bed reactor nuclear rocket concept consists of fuel particles (in this case (U,Zr)C with an outer coat of zirconium carbide). These particles are packed in an annular bed surrounded by two frits (porous tubes) forming a fuel element; the outer one being a cold frit, the inner one being a hot frit. The fuel element are cooled by hydrogen passing in through the moderator. These elements are assembled in a reactor assembly in a hexagonal pattern. The reactor can be either reflected or not, depending on the design, and either 19 or 37 elements, are used. Propellant enters in the top, passes through the moderator fuel element and out through the nozzle. Beryllium used for the moderator in this particular design to withstand the high radiation exposure implied by the long run times

40 CFR 471.90 - Applicability; description of the zirconium-hafnium forming subcategory.

Science.gov (United States)

2010-07-01

... zirconium-hafnium forming subcategory. 471.90 Section 471.90 Protection of Environment ENVIRONMENTAL... POINT SOURCE CATEGORY Zirconium-Hafnium Forming Subcategory § 471.90 Applicability; description of the zirconium-hafnium forming subcategory. This subpart applies to discharges of pollutants to waters of the...

Device and method for shortening reactor process tubes

Science.gov (United States)

Frantz, Charles E.; Alexander, William K.; Lander, Walter E. B.

1980-01-01

This disclosure describes a device and method for in situ shortening of nuclear reactor zirconium alloy process tubes which have grown as a result of radiation exposure. An upsetting technique is utilized which involves inductively heating a short band of a process tube with simultaneous application of an axial load sufficient to cause upsetting with an attendant decrease in length of the process tube.

Influence of zirconium ions on the sorption of carrier-free radiophosphate (32P)

International Nuclear Information System (INIS)

Friedmann, Ch.; Schoenfeld, T.

1975-01-01

In acid solutions the addition of zirconium ions largely affects the sorption of carrier-free radiophosphate on various materials. With some sorbents, such as diatomeceous earth, clay minerals or activated charcoal, the addition of small quantities of zirconium leads to a substantial increase of 32 P adsorption. On the other hand, important quantities of zirconium cause decrease of sorption. With alumina as an adsorbent, any addition of zirconium leads to reduced adsorption of radiophosphate. These phenomena are due to the formation of soluble zirconium-phosphate complex ions. (author)

Nuclear calculation for employing medium enrichment in reactors of Japan Atomic Energy Research Institute

International Nuclear Information System (INIS)

Miyasaka, Yasuhiko

1979-01-01

The fuel used for the research reactors of Japan Atomic Energy Research Institute (JAERI) is presently highly enriched uranium of 93%. However, the U.S. government (the supplier of fuel) is claiming to utilize low or medium enriched uranium from the viewpoint of resistivity to nuclear proliferation, and the availability of highly enriched uranium is becoming hard owing to the required procedure. This report is described on the results of nuclear calculation which is the basis of fuel design in the countermeasures to the reduction of enrichment. The basic conception in the reduction of enrichment is three-fold: to lower the latent potential of nuclear proliferation as far as possible, to hold the present reactor performance as far as possible, and to limit the reduction in the range which is not accompanied by the modification of reactor core construction and cooling system. This time, the increase of the density and thickness of fuel plates and the effect of enrichment change to 45% on reactivity and neutron flux were investigated. The fuel of UAl sub(x) - Al system was assumed, which was produced by powder metallurgical method. The results of investigations on JRR-2 and JMTR reactors revealed that 45% enriched fuel does not affect the performances much. However, deterioration of the performances is not neglegible if further reduction is needed. In future, the influence of the burn-up effect of fuel on the life of reactor cores must be investigated. (Wakatsuki, Y.)

Photometric determination of zirconium in phosphorites by reaction with arsenazo III

Energy Technology Data Exchange (ETDEWEB)

Nikol' skaya, I V; Maksimov, A V

1976-05-01

The reaction between zirconium and arsenazo III has been studied over a wide range of hydrochloric acid concentration and under different conditions. 6 and 9 M HCl solutions are optimal for determining zirconium; the least effect of phosphate ions and color stability in time are observed in this case. The determination of zirconium should be carried out using 10-fold reagent excess and in 15-20 min after adding the reagent. The interference of phosphate ions has been estimated. A procedure has been developed for photometric determination of zirconium in phosphorites with prior acid separation of soluble impurities.

Characterization of the insoluble sludge from the dissolution of irradiated fast breeder reactor fuel

Energy Technology Data Exchange (ETDEWEB)

Aihara, Haruka; Arai, Yoichi; Shibata, Atsuhiro; Nomura, K.; Takeuchi, M. [Japan Atomic Energy Agency - JAEA, 4-33 Muramatsu, Tokai-mura, Naka-gun, Ibaraki, 319-1194 (Japan)

2016-07-01

Insoluble sludge is generated in the reprocessing of spent fuel. The sludge obtained from the dissolution of irradiated fuel from the Joyo experimental fast reactor was analyzed to evaluate its chemical form. The sludge was collected by the filtration of the dissolved fuel solution, and then washed in nitric acid. The yields of the sludge weight were less than 1% of the total fuel weight. The chemical composition of the sludge was analyzed after decomposition by alkaline fusion. Molybdenum, technetium, ruthenium, rhodium, and palladium were found to be the main constituent elements of the sludge. X-ray diffraction patterns of the sludge were attributable to Mo{sub 4}Ru{sub 4}RhPd, regardless of the experimental conditions. The concentrations of molybdenum and zirconium in the dissolved fast reactor fuel solutions were low, indicating that zirconium molybdate hydrate (ZMH) is produced in negligible amounts in the process. (authors)

Preparation of complexes of zirconium and hafnium tetrachlorides with phosphorus oxychloride

International Nuclear Information System (INIS)

Snyder, T.S.; Stoltz, R.A.

1989-01-01

This patent describes an improvement in a method for separating hafnium chloride from zirconium chloride using a distillation column, with a hafnium chloride enriched vapor stream taken from the top of the column and a zirconium enriched chloride stream taken from the bottom of the column. The improvement comprising: purifying the zirconium-hafnium chloride in a molten salt purification vessel prior to or after introduction into the distillation column to substantially eliminate iron chloride from the zirconium-hafnium chloride by at least periodically removing iron chloride from the molten salt purification vessel by electrochemically plating iron onto an electrode in the molten salt purification vessel. The molten salt in the molten salt purification vessel consisting essentially of a mixture of chlorides selected from the group consisting of alkali metals, alkaline earth metals, zirconium, hafnium, aluminum, manganese, and zinc

Zirconium distribution in the system HNO3-H2O-TBP-diluent

International Nuclear Information System (INIS)

Shu, J.; Araujo, B.F. de.

1984-01-01

The extraction behaviour of zirconium in TBP/diluent-HNO 3 -H 2 O systems is studied in order to increase the uranium decontamination factor by adjusting the extraction conditions so that zirconium extraction is kept at a minimum. Equilibrium diagram, TBP concentration, aqueous: organic phases ratio, salting-out effects and uranium loading in the organic phase were the main factors studied. All the experiments have been carried out with zirconium in the 10 -2 - 10 -3 M concentration range. The extractant degradation products influence upon zirconium behaviour was also verified. With the data obtained it was possible to introduce some modifications in the standard Purex flowsheet with the increase of the decontamination of uranium product from zirconium. (Author) [pt

Determination of uranium traces in nuclear cans of nuclear reactors

International Nuclear Information System (INIS)

Acosta L, E.; Benavides M, A.M.; Sanchez P, L.

1996-01-01

To quantify the uranium content as impurity can be found in zirconium alloys and zircaloy, utilized to construct the sheaths containing fuels of the reactors of nuclear plants. The determination by fluorescence spectroscopy was employed as quality control measurement, at once the corrosion resistance, diminish with the increase of the uranium content in the alloys. (Author)

Reactor technology

International Nuclear Information System (INIS)

Erdoes, P.

1977-01-01

This is one of a series of articles discussing aspects of nuclear engineering ranging from a survey of various reactor types for static and mobile use to mention of atomic thermo-electric batteries of atomic thermo-electric batteries for cardiac pacemakers. Various statistics are presented on power generation in Europe and U.S.A. and economics are discussed in some detail. Molten salt reactors and research machines are also described. (G.M.E.)

Sorbents based on xerogels of zirconium, aluminum and manganese oxyhydroxides

Directory of Open Access Journals (Sweden)

R.V. Smotraiev

2016-05-01

Full Text Available The actual problem of water supply in the world and in Ukraine, in particular, is a high level of pollution in water resources and an insufficient level of drinking water purification. With industrial wastewater, a significant amount of pollutants falls into water bodies, including suspended particles, sulfates, iron compounds, heavy metals, etc. Aim: The aim of this work is to determine the impact of aluminum and manganese ions additives on surface and sorption properties of zirconium oxyhydroxide based sorbents during their production process. Materials and Methods: The sorbents based on xerogels of zirconium, aluminum and manganese oxyhydroxides were prepared by sol-gel method during the hydrolysis of metal chlorides (zirconium oxychloride ZrOCl2, aluminum chloride AlCl3 and manganese chloride MnCl2 with carbamide. Results: The surface and sorption properties of sorbents based on xerogels of zirconium, aluminum and manganese oxyhydroxides were investigated. X-ray amorphous structure and evolved hydroxyl-hydrate cover mainly characterize the obtained xerogels. The composite sorbents based on xerogels of zirconium oxyhydroxide doped with aluminum oxyhydroxide (aS = 537 m2/g and manganese oxyhydroxide (aS = 356 m2/g have more developed specific surface area than single-component xerogels of zirconium oxyhydroxide (aS = 236 m2/g and aluminum oxyhydroxide (aS = 327 m2/g. The sorbent based on the xerogel of zirconium and manganese oxyhydroxides have the maximum SO42--ions sorption capacity. It absorbs 1.5 times more SO42–-ions than the industrial anion exchanger AN-221. The sorbents based on xerogels of zirconium oxyhydroxide has the sorption capacity of Fe3+-ions that is 1.5…2 times greater than the capacity of the industrial cation exchanger KU-2-8. The Na+-ions absorption capacity is 1.47…1.56 mmol/g for each sorbent. Conclusions: Based on these data it can be concluded that the proposed method is effective for sorbents production based on

Modification of zirconium diphosphate with salicylic acid and its effect on the uranium (Vi) sorption; Modificacion del difosfato de circonio con acido salicilico y su efecto sobre la sorcion de uranio (VI)

Energy Technology Data Exchange (ETDEWEB)

Almazan T, M. G.; Garcia G, N. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Simoni, E., E-mail: guadalupe.almazan@inin.gob.mx [Universidad Paris Sud, Instituto de Fisica Nuclear, Georges Clemenceau No. 15, Orsay (France)

2014-10-15

The surface of zirconium diphosphate (ZrP{sub 2}O{sub 7}) was modified with salicylic acid and its effect was evaluated on the uranium (Vi) sorption. The modified surface of the material was analyzed with different analytical techniques among which are included the atomic force microscopy, scanning electron microscopy and X-ray photoelectron spectroscopy. This analysis allowed showing that the salicylic acid is being held on the surface of the zirconium diphosphate. The reactivity of modified zirconium diphosphate compared with uranium (Vi) was investigated using the classical method of batch sorption. The analysis of sorption isotherms shows that the salicylic acid has an important effect in the uranium (Vi) sorption. According to the study conducted, the interaction among the uranium (Vi) and the surface of zirconium diphosphate modified with the salicylic acid most likely leads to the complexes formation of binary (U(Vi)/ZrP{sub 2}O{sub 7}) and ternary (U(Vi)/salicylate/ZrP{sub 2}O{sub 7}) surface. (Author)

Manufacturing process to reduce large grain growth in zirconium alloys

International Nuclear Information System (INIS)

Rosecrans, P.M.

1987-01-01

A method is described of treating cold worked zirconium alloys to reduce large grain growth during thermal treatment above its recrystallization temperature. The method comprises heating the zirconium alloy at a temperature of about 1300 0 F. to 1350 0 F. for about 1 to 3 hours subsequent to cold working the zirconium alloy and prior to the thermal treatment at a temperature of between 1450 0 -1550 0 F., the thermal treatment temperature being above the recrystallization temperature

Properties of zirconium carbide for nuclear fuel applications

Energy Technology Data Exchange (ETDEWEB)

Katoh, Yutai; Vasudevamurthy, Gokul, E-mail: gvasudev@vcu.edu; Nozawa, Takashi; Snead, Lance L.

2013-10-15

Zirconium carbide (ZrC) is a potential coating, oxygen-gettering, or inert matrix material for advanced high temperature reactor fuels. ZrC has demonstrated attractive properties for these fuel applications including excellent resistance against fission product corrosion and fission product retention capabilities. However, fabrication of ZrC results in a range of stable sub-stoichiometric and carbon-rich compositions with or without substantial microstructural inhomogeneity, textural anisotropy, and a phase separation, leading to variations in physical, chemical, thermal, and mechanical properties. The effects of neutron irradiation at elevated temperatures, currently only poorly understood, are believed to be substantially influenced by those compositional and microstructural features further adding complexity to understanding the key ZrC properties. This article provides a survey of properties data for ZrC, as required by the United States Department of Energy’s advanced fuel programs in support of the current efforts toward fuel performance modeling and providing guidance for future research on ZrC for fuel applications.

Measurement and analysis of pressure tube elongation in the Douglas Point reactor

International Nuclear Information System (INIS)

Causey, A.R.; MacEwan, S.R.; Jamieson, H.C.; Mitchell, A.B.

1980-02-01

Elongations of zirconium alloy pressure tubes in CANDU reactors, which occur as a result of neutron-irradiation-induced creep and growth, have been measured over the past 6 years, and the consequences of thses elongations have recently been analysed. Elongation rates, previously deduced from extensive measurements of elongations of cold-worked Zircaloy-2 pressure tubes in the Pickering reactors, have been modified to apply to the pressure tubes in the Douglas Point (DP) reactor by taking into account measured diffences in texture and dislocation density. Using these elongation rates, and structural data unique to the DP reactor, the analysis predicts elongation behaviour which is in good agreement with pressure tube elongations measured during the ten years of reactor operation. (Auth)

Atoms at work

International Nuclear Information System (INIS)

1982-07-01

This illustrated booklet discusses the following: atoms; fission of uranium; nuclear power plants; reactor types; plutonium (formation, properties, uses); radioactive waste (fuel cycle, reprocessing, waste management); nuclear fusion; fusion reactors; radiation; radioisotopes and their uses. (U.K.)

Method of purifying zirconium tetrachloride and hafnium tetrachloride in a vapor stream

International Nuclear Information System (INIS)

Snyder, T.S.; Stolz, R.A.

1992-01-01

This patent describes a method of purifying zirconium tetrachloride and hafnium tetrachloride in a vapor stream from a sand chlorinator in which the silicon and metals present in sand fed to the chlorinator are converted to chlorides at temperatures over about 800 degrees C. It comprises cooling a vapor stream from a sand chlorinator, the vapor stream containing principally silicon tetrachloride, zirconium tetrachloride, and hafnium tetrachloride contaminated with ferric chloride, to a temperature of from about 335 degrees C to about 600 degrees C; flowing the vapor stream through a gaseous diffusion separative barrier to produce a silicon tetrachloride-containing vapor stream concentrated in zirconium tetrachloride and hafnium tetrachloride and a silicon tetrachloride-containing vapor stream depleted in zirconium tetrachloride and hafnium tetrachloride; adsorbing the ferric chloride in the separative barrier; and recovering the silicon tetrachloride stream concentrated in zirconium tetrachloride and hafnium tetrachloride separately from the silicon tetrachloride stream depleted in zirconium tetrachloride and hafnium tetrachloride

Determination of impurities in zirconium by emission spectrograph method

International Nuclear Information System (INIS)

Simbolon, S.; Masduki, B.; Aryadi

2000-01-01

Analysis of B, Cd, Si and Cr elements in zirconium oxide was carried out. Zirconium oxide was made by precipitating zirconium solution with oxalic acid and calcination was at temperature 900 oC for four hours. Silver chloride compound as much as 10% was used as a distillation carrier and 7 step filtration was used to reduce the impurities element spectra having high density. It was found that B concentration is between 3.80 and 7.44 ppm, Cd less then 0.5 ppm, Si between 74.38-150.33 ppm and Cr between 19.90-45.76 ppm. (author)

Preconcentration of uranium, thorium, zirconium, titanium, molybdenum and vanadium with oxine supported on microcrystalline naphthalene and their determinations by inductively coupled plasma atomic emission spectrometry

International Nuclear Information System (INIS)

Naveen Kumar, P.; Sanjay Kumar; Vijay Kumar; Nandakishore, S.S.; Bangroo, P.N.

2013-01-01

A sensitive and rapid method for the determination of uranium, thorium, zirconium, titanium, molybdenum and vanadium by inductively coupled plasma atomic emission spectrometry (ICP-AES) after solid-liquid extraction with microcrystalline naphthalene is developed. Analytes were quantitatively adsorbed as their oxinate complexes on naphthalene and determined by ICP-AES after stripping with 2 M HCl. The effect of various experimental parameters such as pH, reagent amounts, naphthalene amount and stripping conditions on the determination of these elements was investigated in detail. Under the optimized experimental conditions, the detection limits of this method for U (VI), Th (IV), Zr (IV), Ti (IV), Mo (VI) and V (V) were 20.0 ng mL -1 and the relative standard deviations obtained for three replicate determinations at a concentration of 1.0 µg mL -1 were 1.5-3.0%. The proposed method has been applied in the analysis of SY-2, SY-3 and pre-analysed samples for U, Th, Zr, Ti, Mo and V the analytical results are in good agreement with recommended values. (author)

Fluorometric determination of zirconium in minerals

Science.gov (United States)

Alford, W.C.; Shapiro, L.; White, C.E.

1951-01-01

The increasing use of zirconium in alloys and in the ceramics industry has created renewed interest in methods for its determination. It is a common constituent of many minerals, but is usually present in very small amounts. Published methods tend to be tedious, time-consuming, and uncertain as to accuracy. A new fluorometric procedure, which overcomes these objections to a large extent, is based on the blue fluorescence given by zirconium and flavonol in sulfuric acid solution. Hafnium is the only element that interferes. The sample is fused with borax glass and sodium carbonate and extracted with water. The residue is dissolved in sulfuric acid, made alkaline with sodium hydroxide to separate aluminum, and filtered. The precipitate is dissolved in sulfuric acid and electrolysed in a Melaven cell to remove iron. Flavonol is then added and the fluorescence intensity is measured with a photo-fluorometer. Analysis of seven standard mineral samples shows excellent results. The method is especially useful for minerals containing less than 0.25% zirconium oxide.

Low stress creep behaviour of zirconium

International Nuclear Information System (INIS)

Prasad, N.

1989-01-01

Creep behaviour of alpha zirconium of grain size varying between 16 and 55 μm has been investigated in the temperature range 813 to 1003K at stresses upto 5.5 MNm -2 using high sensitive spring specimen geometry. Creep experiments on specimens of 50 μm grain size revealed a transition from lattice diffusion controlled viscous creep at temperatures greater than 940K to grain boundary diffusion controlled viscous creep at lower temperatures. Tests conducted on either side of the transition suggest the dominance of Nabarro-Herring and Coble creep processes respectively. Evidence for power-law creep has been observed in practically all the creep tests. Based on the experimental data obtained in the present study and those recently reported by Novotny et al (1985), Langdon creep mechanism maps have bee n constructed at 873 and 973K. With the help of these maps for zirconium and those published for titanium the low stress creep behaviour of zirconium and titanium are compared. (author). 22 refs., 11 figs., 3 tabs

Electrocnecical behaviour of zirconium during its anodic polarization in nitrate solutions

International Nuclear Information System (INIS)

Stabrovskij, A.I.; Karasev, A.F.

1983-01-01

Electrochemical behaviour of zirconium during its anodic polarization in nitrate solutions is investigated in detail to find the method of its complete dissolution. A study has been made of the influence of varioUs factors: current density electric potential, composition and temperature of the solution, anodic polarization duration on the Zr anodic polarization in nitric acid, on the maximum permissible current density and on the zirconium yield to the solution. The zirconium polarization decreases with an acid concentration and temperature increase and increases with the current density. Iron nitrate additions to nitric acid decrease, while ammonium fluoride additions increase zirconium yield into the solution

Production optimization of {sup 99}Mo/{sup 99m}Tc zirconium molybate gel generators at semi-automatic device: DISIGEG

Energy Technology Data Exchange (ETDEWEB)

Monroy-Guzman, F., E-mail: fabiola.monroy@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera Mexico-Toluca S/N, La Marquesa, Ocoyoacac, 52750, Estado de Mexico (Mexico); Rivero Gutierrez, T., E-mail: tonatiuh.rivero@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera Mexico-Toluca S/N, La Marquesa, Ocoyoacac, 52750, Estado de Mexico (Mexico); Lopez Malpica, I.Z.; Hernandez Cortes, S.; Rojas Nava, P.; Vazquez Maldonado, J.C. [Instituto Nacional de Investigaciones Nucleares, Carretera Mexico-Toluca S/N, La Marquesa, Ocoyoacac, 52750, Estado de Mexico (Mexico); Vazquez, A. [Instituto Mexicano del Petroleo, Eje Central Norte Lazaro Cardenas 152, Col. San Bartolo Atepehuacan, 07730, Mexico D.F. (Mexico)

2012-01-15

DISIGEG is a synthesis installation of zirconium {sup 99}Mo-molybdate gels for {sup 99}Mo/{sup 99m}Tc generator production, which has been designed, built and installed at the ININ. The device consists of a synthesis reactor and five systems controlled via keyboard: (1) raw material access, (2) chemical air stirring, (3) gel dried by air and infrared heating, (4) moisture removal and (5) gel extraction. DISIGEG operation is described and dried condition effects of zirconium {sup 99}Mo- molybdate gels on {sup 99}Mo/{sup 99m}Tc generator performance were evaluated as well as some physical-chemical properties of these gels. The results reveal that temperature, time and air flow applied during the drying process directly affects zirconium {sup 99}Mo-molybdate gel generator performance. All gels prepared have a similar chemical structure probably constituted by three-dimensional network, based on zirconium pentagonal bipyramids and molybdenum octahedral. Basic structural variations cause a change in gel porosity and permeability, favouring or inhibiting {sup 99m}TcO{sub 4}{sup -} diffusion into the matrix. The {sup 99m}TcO{sub 4}{sup -} eluates produced by {sup 99}Mo/{sup 99m}Tc zirconium {sup 99}Mo-molybdate gel generators prepared in DISIGEG, air dried at 80 Degree-Sign C for 5 h and using an air flow of 90 mm, satisfied all the Pharmacopoeias regulations: {sup 99m}Tc yield between 70-75%, {sup 99}Mo breakthrough less than 3 Multiplication-Sign 10{sup -3}%, radiochemical purities about 97% sterile and pyrogen-free eluates with a pH of 6. - Highlights: Black-Right-Pointing-Pointer {sup 99}Mo/{sup 99m}Tc generators based on {sup 99}Mo-molybdate gels were synthesized at a semi-automatic device. Black-Right-Pointing-Pointer Generator performances depend on synthesis conditions of the zirconium {sup 99}Mo-molybdate gel. Black-Right-Pointing-Pointer {sup 99m}TcO{sub 4}{sup -} diffusion and yield into generator depends on gel porosity and permeability. Black

Mechanical, electronic, and thermodynamic properties of zirconium carbide from first-principles calculations

Science.gov (United States)

Yang, Xiao-Yong; Lu, Yong; Zheng, Fa-Wei; Zhang, Ping

2015-11-01

Mechanical, electronic, and thermodynamic properties of zirconium carbide have been systematically studied using the ab initio calculations. The calculated equilibrium lattice parameter, bulk modulus, and elastic constants are all well consistent with the experimental data. The electronic band structure indicates that the mixture of C 2p and Zr 4d and 4p orbitals around the Fermi level makes a large covalent contribution to the chemical bonds between the C and Zr atoms. The Bader charge analysis suggests that there are about 1.71 electrons transferred from each Zr atom to its nearest C atom. Therefore, the Zr-C bond displays a mixed ionic/covalent character. The calculated phonon dispersions of ZrC are stable, coinciding with the experimental measurement. A drastic expansion in the volume of ZrC is seen with increasing temperature, while the bulk modulus decreases linearly. Based on the calculated phonon dispersion curves and within the quasi-harmonic approximation, the temperature dependence of the heat capacities is obtained, which gives a good description compared with the available experimental data. Project supported by the National Natural Science Foundation of China (Grant No. 51071032).

Run-Beyond-Cladding-Breach (RBCB) test results for the Integral Fast Reactor (IFR) metallic fuels program

International Nuclear Information System (INIS)

Batte, G.L.; Hoffman, G.L.

1990-01-01

In 1984 Argonne National Laboratory (ANL) began an aggressive program of research and development based on the concept of a closed system for fast-reactor power generation and on-site fuel reprocessing, exclusively designed around the use of metallic fuel. This is the Integral Fast Reactor (IFR). Although the Experimental Breeder Reactor-II (EBR-II) has used metallic fuel since its creation 25 yeas ago, in 1985 ANL began a study of the characteristics and behavior of an advanced-design metallic fuel based on uranium-zirconium (U-Zr) and uranium-plutonium-zirconium (U-Pu-Zr) alloys. During the past five years several areas were addressed concerning the performance of this fuel system. In all instances of testing the metallic fuel has demonstrated its ability to perform reliably to high burnups under varying design conditions. This paper will present one area of testing which concerns the fuel system's performance under breach conditions. It is the purpose of this paper to document the observed post-breach behavior of this advanced-design metallic fuel. 2 figs., 1 tab

Spectrofluorimetric determination of hafnium and zirconium with 3,7-dihydroxyflavone

Energy Technology Data Exchange (ETDEWEB)

Ito, Takushi; Suzuki, Osamu; Seuzuki, Tetsuo; Murata, Akira

1986-04-01

The absorptive and fluorescent characteristics of the hafnium and zirconium complexes of 3-hydroxyflavone and its 12 hydroxy and methoxy derivatives have been studied. The fluorescence of the 1:1 hafnium - 3,7-dihydroxyflavone complex (lambdasub(ex.)397 nm,lambdasub(em.) 465 nm) in 3 M hydrochloric acid has been used to determine 2-40 ng ml/sup -1/ of hafnium. The fluorescence of the 1:1 zirconium - 3,7-dihydroxyflavone complex (lambdasub(ex.) 395 nm, lambda sub(em.) 465 nm) at pH 2.0 in 0.02 M sulphate solution has been used to determine 2-40 ng ml/sup -1/ of zirconium. These methods are very sensitive and can be used for the simultaneous determination of hafnium and zirconium with an error of about 5%.

Research into zirconium alloys resistant to carbon dioxide under pressure at temperatures of up to 600 deg C (1963)

International Nuclear Information System (INIS)

Baque, P.; Dominget, R.; Bossard, J.

1963-01-01

Zirconium is a metal having a relatively low neutron capture cross-section and a high melting point; it is thus possible to consider its use in particular as a canning material for fuel elements in CO 2 -cooled nuclear reactors. A preliminary study of several types of zirconium showed that the metal is already strongly oxidised in this gas at 500 deg C. The 'breakaway' phenomenon is generalised; the oxidation rate is then linear and depends on the carbon dioxide pressure. An attempt was therefore made to find binary and tertiary alloys in order to improve the metal behaviour. Several interesting compositions were found: 1, 1.6 and 2.5 per cent of copper, 2 per cent of vanadium, and 0.05 and 0.5 per cent of calcium. Tertiary copper-molybdenum and copper-phosphorus alloys are also less liable to oxidation and in particular do not exhibit the 'breakaway' phenomenon even after a prolonged treatment at 600 deg C. (authors) [fr

Geologic structure of Gofitsky deposit of titanium and zirconium and perspectives of the reserve base of titanium and zirconium in Russia

Science.gov (United States)

Kukhmazov, Iskander

2016-04-01

With the fall of the Soviet Union, all the mining deposits of titanium and zirconium appeared outside of Russian Federation. Therefore the studying of deposits of titanium and zirconium in Russia is very important nowadays. There is a paradoxical situation in the country: in spite of possible existence of national mineral resource base of Ti-Zr material, which can cover needs of the country, Russia is the one of the largest buyers of imported Ti-Zr material in the world. Many deposits are not mined, and those which are in the process of mining have poor reserves. Demand for this raw material is very great not only for Russia, but also for the world in general. Today there is a scarcity of zircon around the world and it will only increase through time. Therefore prices of products of titanium and zirconium also increase. Consequently Russian deposits of titanium and zirconium with higher content than foreign may become competitive. Russia is forced to buy raw materials (zirconium and titanium production) from former Soviet Union countries at prices higher than the world's and thus incur huge losses, including customs charges. Russia should create its own mineral resource base of Ti-Zr. Studied titanium-zirconium deposits of Stavropol region may become the basis for the south part of Russia. At first, Beshpagirsky deposit should be pointed out. It has large reserves of ore sands with high content of Ti-Zr. A combination of favorable geographical position of the area with developed industrial infrastructure makes it very beneficial as an object for high priority development. Gofitsky deposit should be pointed out as well. Its sands have a wide areal distribution and a high content of titanium and zirconium. Chokrak, Karagan-Konksk and Sarmatian sediments of the Miocene of Gofitsky deposit are productive for titanium and zirconium placers within Stavropol region of Russia. Gofitsky deposit was evaluated from financial and economic point of view and the following data

Collaborative study of the colorimetric determination of zirconium in antiperspirant aerosols

International Nuclear Information System (INIS)

Beavin, P. Jr.

1977-01-01

A previously published method for determining zirconium in antiperspirant aerosols was collaboratively studied by 7 laboratories. The method consists of 2 procedures: a rapid dilution procedure for soluble zirconium compounds or a lengthier fusion procedure for total zirconium followed by colorimetric determination. The collaborators were asked to perform the following: Spiking materials representing 4 levels of soluble zirconium were added to weighed portions of a zirconium-free cream base concentrate and the portions were assayed by the dilution procedure. Spiking materials representing 4 levels of zirconium in either the soluble or the insoluble form (or as a mixture) were also added to portions of the same concentrate and these portions were assayed by the fusion procedure. They were also asked to concentrate and assay, by both procedures, 2 cans each of 2 commercial aerosol antiperspirants containing zirconyl hydroxychloride. The average percent recoveries and standard deviations for spiked samples were 99.8-100.2 and 1.69-2.71, respectively, for soluble compounds determined by the dilution procedure, and 93.8-97.4 and 3.09-4.78, respectively, for soluble and/or insoluble compounds determined by the fusion procedure. The average perent zirconium found by the dilution procedure in the 2 commercial aerosol products was 0.751 and 0.792. Insufficient collaborative results were received for the fusion procedure for statistical evaluation. The dilution procedure has been adopted as official first action

Corrosion resistance and biocompatibility of zirconium oxynitride thin film growth by RF sputtering

Energy Technology Data Exchange (ETDEWEB)

Cubillos, G. I.; Olaya, J. J.; Clavijo, D.; Alfonso, J. E. [Universidad Nacional de Colombia, Carrera 45 No. 26-85, AA 14490 Bogota D. C. (Colombia); Bethencourt, M., E-mail: jealfonsoo@unal.edu.co [Universidad de Cadiz, Centro Andaluz de Ciencia y Tecnologia Marinas, Departamento de Ciencia de los Materiales e Ingenieria Metalurgica y Quimica Inorganica, Av. Republica de Saharaui, Puerto Real, E-11510 Cadiz (Spain)

2012-07-01

Thin films of zirconium oxynitride were grown on common glass, silicon (100) and stainless steel 316 L substrates using the reactive RF magnetron sputtering technique. The films were analyzed through structural, morphological and biocompatibility studies. The structural analysis was carried out using X-ray diffraction (XRD), and the morphological analysis was carried out using scanning electron microscopy (Sem) and atomic force microscopy (AFM). These studies were done as a function of growth parameters, such as power applied to the target, substrate temperature, and flow ratios. The corrosion resistance studies were made on samples of stainless steel 316 L coated and uncoated with Zr{sub x}N{sub y}O films, through of polarization curves. The studies of biocompatibility were carried out on zirconium oxynitride films deposited on stainless steel 316 L through proliferation and cellular adhesion. The XRD analysis shows that films deposited at 623 K, with a flow ratio {Phi}N{sub 2}/{Phi}O{sub 2} of 1.25 and a total deposit time of 30 minutes grew preferentially oriented along the (111) plane of the zirconium oxynitride monoclinic phase. The Sem analyses showed that the films grew homogeneously, and the AFM studies indicated that the average rugosity of the film was 5.9 nm and the average particle size was 150 nm. The analysis of the corrosion resistant, shows that the stainless steel coated with the film was increased a factor 10. Finally; through the analysis of the biocompatibility we established that the films have a better surface than the substrate (stainless steel 316 L) in terms of the adhesion and proliferation of bone cells. (Author)

Corrosion of zirconium alloys in nuclear reactors: A model for irradiation induced enhancement by local radiolysis in the porous oxide

Energy Technology Data Exchange (ETDEWEB)

Lemaignan, C; Salot, R [CEA/DRN/DTP, CENG-SECC, Grenoble (France)

1997-02-01

An analysis has been undertaken of the various cases of local enhancement of corrosion rate of zirconium alloys under irradiation. It is observed that in most cases a strong emission of energetic {beta}{sup -} is present leading to a local energy deposition rate higher than the core average. This suggests that the local transient radiolytic oxidizing species produced in the coolant by the {beta}{sup -} particles could contribute to corrosion enhancement, by increasing the local corrosion potential. This process is applicable to the local enhanced corrosion found in front of stainless steels structural parts, due to the contribution of Mn, and in front of Pt inserts or Cu-rich cruds. It explains also the irradiation corrosion enhancement of Cu-Zr alloys. Enhanced corrosion around neutron absorbing material is explained similarly by pair production from conversion of high energy capture photons in the cladding, leading to energetic electrons. The same process was found to be active with other highly ionizing species like {alpha} from Ni-rich alloys and fission products in homogeneous reactors. Due to the changes induced by the irradiation intensity on the concentration of the radiolytic species, the coolant chemistry, that controls the boundary conditions for oxide growth, has to be analyzed with respect to the local value of the energy deposition rate. An analysis has been undertaken which shows that, in a porous media, the water is exposed to a higher intensity than bulk water. This leads to a higher concentration of oxidizing radiolytic species at the root of the cracks of the porous oxide, and increases the corrosion rate under irradiation. This mechanism, deduced from the explanation proposed for localized irradiation enhanced corrosion, can be extended to the whole reactor core, where the general enhancement of Zr alloys corrosion under irradiation could be attributed to the general radiolysis in the porous zirconia. (author). 18 refs, 3 figs, 3 tabs.

Hydrogen outbreak of Zirconium Molybdate Hihydrate

International Nuclear Information System (INIS)

Miura, Yasuhiko; Fukuda, Kazuhiro; Ochi, Eiji

2008-01-01

JNFL is planning to construct a facility for enclosing the hull and end pieces produced due to reprocessing of spent fuel into stainless canisters after compressing, while those hull and end pieces enclosed into the stainless canisters are called 'compressed hulls'. Since the compressed hulls contain moisture absorbent Zirconium Molybdate Hihydrate accompanying hull and end pieces, there is a risk of outbreak of radiolysisradiolysis gas such as hydrogen, etc. by radiolysisradiolysis. This report intends to state the result of radiation irradiation experiment with the purpose of examining the volume of hydrogen outbreak from Zirconium Molybdate Hihydrate of the compressed hulls. (author)

Evaluation and optimization of General Atomics' GT-MHR reactor cavity cooling system using an axiomatic design approach

International Nuclear Information System (INIS)

Thielman, Jeff; Ge, Ping; Wu, Qiao; Parme, Laurence

2005-01-01

The development of the Generation IV (Gen-IV) nuclear reactors has presented social, technical, and economical challenges to nuclear engineering design and research. To develop a robust, reliable nuclear reactor system with minimal environmental impact and cost, modularity has been gradually accepted as a key concept in designing high-quality nuclear reactor systems. While the establishment and reliability of a nuclear power plant is largely facilitated by the installment of standardized base units, the realization of modularity at the sub-system/sub-unit level in a base unit is still highly heuristic, and lacks consistent, quantifiable measures. In this work, an axiomatic design approach is developed to evaluate and optimize the reactor cavity cooling system (RCCS) of General Atomics' Gas Turbine-Modular Helium Reactor (GT-MHR) nuclear reactor, for the purpose of constructing a quantitative tool that is applicable to Gen-IV systems. According to Suh's axiomatic design theory, modularity is consistently represented by functional independence through the design process. Both qualitative and quantitative measures are developed here to evaluate the modularity of the current RCCS design. Optimization techniques are also used to improve the modularity at both conceptual and parametric level. The preliminary results of this study have demonstrated that the axiomatic design approach has great potential in enhancing modular design, and generating more robust, safer, and less expensive nuclear reactor sub-units

A review of the inorganic and organometallic chemistry of zirconium

International Nuclear Information System (INIS)

Kalvins, A.K.

1985-01-01

The results of a literature review of the inorganic and organometallic chemistry of zirconium are presented. Compounds with physical and chemical properties compatible with the requirements of an ir laser zirconium isotope separation process have been identified

Obtaining of total and thermal neutron flux in the carousel facility of the TRIGA MARK IPR-R1 reactor using the Monte Carlo transport method

International Nuclear Information System (INIS)

Guerra, Bruno Teixeira

2011-01-01

The IPR-R1 is a reactor type TRIGA, Mark-I model, manufactured by the General Atomic Company and installed at Nuclear Technology Development Centre (CDTN) of Brazilian Nuclear Energy Commission (CNEN), in Belo Horizonte, Brazil. It is a light water moderated and cooled, graphite-reflected, open-pool type research reactor. IPR-R1 works at 100 kW but it will be briefly licensed to operate at 250 kW. It presents low power, low pressure, for application in research, training and radioisotopes production. The fuel is an alloy of zirconium hydride and uranium enriched at 20% in 235 U. The goal this work is modelling of the IPR-R1 Research Reactor TRIGA using the codes MCNPX2.6.0 (Monte Carlo N-Particle Transport extend) and MCNP5 to the calculating the neutron flux in the carousel facility. In each simulation the sample was placed in a different position, totaling forty positions around of the reactor core. The comparison between the results obtained with experimental values from other work showing a relatively good agreement. Moreover, this methodology is a theoretical tool in validating of the experimental values and necessary for determining neutron flux which can not be accessible experimentally. (author)

The evolution of Canadian research reactors: 1942 to 1992

International Nuclear Information System (INIS)

Lidstone, R.F.

1992-01-01

Research reactors have played a central role in the development of the Canadian nuclear program. Major benefits arising from the operation of Canada's research reactors include the development and demonstration of Canada deuterium uranium (CANDU) power reactors and the creation of an international business based on the production of medical and industrial radioisotopes. Insofar as practicable, nuclear engineers have based each design on the evolutionary development of appropriate technology for Canada. This approach has resulted in several generations of multipurpose reactors with a common heritage of neutronic efficiency. A focus on neutronic efficiency stemmed from the incentive for direct use of Canada's abundant uranium resources; the desired reliance on natural-uranium fuels promoted interest in efficient moderators, especially heavy water, and structural materials with a low affinity for absorbing neutrons, such as aluminum and zirconium alloys

Research Update: Atmospheric pressure spatial atomic layer deposition of ZnO thin films: Reactors, doping, and devices

Energy Technology Data Exchange (ETDEWEB)

Hoye, Robert L. Z., E-mail: rlzh2@cam.ac.uk, E-mail: jld35@cam.ac.uk; MacManus-Driscoll, Judith L., E-mail: rlzh2@cam.ac.uk, E-mail: jld35@cam.ac.uk [Department of Materials Science and Metallurgy, University of Cambridge, 27 Charles Babbage Road, Cambridge CB3 0FS (United Kingdom); Muñoz-Rojas, David [LMGP, University Grenoble-Alpes, CNRS, F-3800 Grenoble (France); Nelson, Shelby F. [Kodak Research Laboratories, Eastman Kodak Company, Rochester, New York 14650 (United States); Illiberi, Andrea; Poodt, Paul [Holst Centre/TNO Thin Film Technology, Eindhoven, 5656 AE (Netherlands); Roozeboom, Fred [Holst Centre/TNO Thin Film Technology, Eindhoven, 5656 AE (Netherlands); Department of Applied Physics, Eindhoven University of Technology, P.O. Box 513, Eindhoven, 5600 MB (Netherlands)

2015-04-01

Atmospheric pressure spatial atomic layer deposition (AP-SALD) has recently emerged as an appealing technique for rapidly producing high quality oxides. Here, we focus on the use of AP-SALD to deposit functional ZnO thin films, particularly on the reactors used, the film properties, and the dopants that have been studied. We highlight how these films are advantageous for the performance of solar cells, organometal halide perovskite light emitting diodes, and thin-film transistors. Future AP-SALD technology will enable the commercial processing of thin films over large areas on a sheet-to-sheet and roll-to-roll basis, with new reactor designs emerging for flexible plastic and paper electronics.

Mechanical resistance of zirconium implant abutments: A review of the literature

Science.gov (United States)

Vaquero-Aguilar, Cristina; Torres-Lagares, Daniel; Jiménez-Melendo, Manuel; Gutiérrez-Pérez, José L.

2012-01-01

The increase of aesthetic demands, together with the successful outcome of current implants, has renewed interest in the search for new materials with enough mechanical properties and better aesthetic qualities than the materials customarily used in implanto-prosthetic rehabilitation. Among these materials, zirconium has been used in different types of implants, including prosthetic abutments. The aim of the present review is to analyse current scientific evidence supporting the use of this material for the above mentioned purposes. We carried out the review of the literature published in the last ten years (2000 through 2010) of in vitro trials of dynamic and static loading of zirconium abutments found in the databases of Medline and Cochrane using the key words zirconium abutment, fracture resistance, fracture strength, cyclic loading. Although we have found a wide variability of values among the different studies, abutments show favourable clinical behaviour for the rehabilitation of single implants in the anterior area. Such variability may be explained by the difficulty to simulate daily mastication under in vitro conditions. The clinical evidence, as found in our study, does not recommend the use of implanto-prosthetic zirconium abutments in the molar area. Key words: Zirconium abutment, zirconium implant abutment, zirconia abutment, fracture resistance, fracture strength, cyclic loading. PMID:22143702

40 CFR 421.330 - Applicability: Description of the primary zirconium and hafnium subcategory.

Science.gov (United States)

2010-07-01

... primary zirconium and hafnium subcategory. 421.330 Section 421.330 Protection of Environment ENVIRONMENTAL... CATEGORY Primary Zirconium and Hafnium Subcategory § 421.330 Applicability: Description of the primary zirconium and hafnium subcategory. The provisions of this subpart are applicable to discharges resulting...

Laves intermetallics in stainless steel-zirconium alloys

International Nuclear Information System (INIS)

Abraham, D.P.; McDeavitt, S.M.; Richardson, J.W. Jr.

1997-01-01

Laves intermetallics have a significant effect on properties of metal waste forms being developed at Argonne National Laboratory. These waste forms are stainless steel-zirconium alloys that will contain radioactive metal isotopes isolated from spent nuclear fuel by electrometallurgical treatment. The baseline waste form composition for stainless steel-clad fuels is stainless steel-15 wt.% zirconium (SS-15Zr). This article presents results of neutron diffraction measurements, heat-treatment studies and mechanical testing on SS-15Zr alloys. The Laves intermetallics in these alloys, labeled Zr(Fe,Cr,Ni) 2+x , have both C36 and C15 crystal structures. A fraction of these intermetallics transform into (Fe,Cr,Ni) 23 Zr 6 during high-temperature annealing; the authors have proposed a mechanism for this transformation. The SS-15Zr alloys show virtually no elongation in uniaxial tension, but exhibit good strength and ductility in compression tests. This article also presents neutron diffraction and microstructural data for a stainless steel-42 wt.% zirconium (SS-42Zr) alloy

Interrelationship between structure and corrosion behaviour of zirconium

Energy Technology Data Exchange (ETDEWEB)

Guenther, T [Bayer A.G., Leverkusen (Germany, F.R.)

1979-05-01

Due to plant failures caused by the break-down of zirconium grade 702 subjected to sulphuric acid the structure and corrosion behaviour of welded and as delivered specimens were tested for various heat treatments. It was shown by structure investigations and electron microprobe analysis that the corrosion behaviour of zirconium (in boiling 65 pct sulphuric acid) is strongly infuenced by the structure, which in its turn is dependent on the grade of purity and the prehistory of the material. Type, amount, and distribution of residual elements or precipitations caused by them are responsible for the corrosion resistance. This is valid particularly for the element iron. The plant failures mentioned here coincided with the examination results. Measures to improve the chemical resistance of pure zirconium subjected to extremely aggressive media were derived.

A layman's guide to radiation-induced deformation processes in zirconium alloys

International Nuclear Information System (INIS)

Dutton, R.

1990-07-01

The fuel channel (comprising a pressure tube and a calandria tube fabricated from zirconium alloys) in a CANDU reactor undergoes shape changes because of radiation-induced deformation. This is a consequence of the microstructural modification arising from radiation damage produced by the fast-neutron flux. This report summarizes our current understanding of the physical processes responsible for the deformation. With the non-specialist reader in mind, the underlying mechanisms are described in a manner that avoids much of the associated technical terminology. Thus, the basic concepts of plasticity in a crystalline material are introduced and related to the various microstructural defects created during irradiation. In particular, the mechanisms of creep (a time-dependent strain activated by an applied stress) and growth (a time-dependent strain occurring in the absence of stress) are discussed in a non-technical language assisted by simple diagrams. Reference is made to both theoretical investigations (avoiding mathematical complexity) and experimental measurements. It is shown how the qualitative and quantitative knowledge can be used to derive a predictive model for reactor designers and operators. The current status of such a model is evaluated and suggestions for future improvements made

Atomic absorption analysis of serial titanium alloys

International Nuclear Information System (INIS)

Gorlova, M.N.; Feofanova, N.M.; Kornyushkova, Yu.D.

1977-01-01

Atom-absorption technique is described, which makes it possible to rapidly and precisely determine the following alloying elements and admixtures in titanium alloys: Al (2.0 - 8.5%); Mo (0.5 - 8%); Cr (0.5 - 12%); Si (0.2 - 0.5%); Mn(0.2 - 2.5%); V(0.5 - 6%); Sn(2.0 - 3.0%); Fe(0.1 - 1.0%); Zr(2.0 - 12.0%). The atom absorption method with flame atomization of the sample provides for best results if the alloy is dissolved in a mixture HCl + HBF 4 in the ratio 2:1. In order to obtain correct results the standard solutions must contain titanium in concentrations corresponding to the weight of the sample being analyzed. Sensitivity of zirconium determination may be increased approximately twofold by adding 10 mg/ml of FeCl 3 into the solution. Being as precise, as the classic analytical methods, the atom absorption technique is about 5 times more efficient

CAC-RA1 1958-1998. The first years of the Constituyentes Atomic Center (CAC). History of the first Argentine nuclear reactor (RA-1)

International Nuclear Information System (INIS)

Forlerer, Elena; Palacios, Tulio A.

1998-01-01

After giving the milestones of the development of the Constituyentes Atomic Center since 1957, the history of the construction of the first nuclear reactor (RA-1) in Argentina, including the local fabrication of its fuel elements, is surveyed. The RA-1 reached criticality on January 17, 1958. The booklet commemorates the 40th year of the reactor operation

Monochloro non-bridged half-metallocene-type zirconium complexes containing phosphine oxide-(thio)phenolate chelating ligands as efficient ethylene polymerization catalysts.

Science.gov (United States)

Tang, Xiao-Yan; Wang, Yong-Xia; Liu, San-Rong; Liu, Jing-Yu; Li, Yue-Sheng

2013-01-14

A series of novel monochloro half-zirconocene complexes containing phosphine oxide-(thio)phenolate chelating ligands of the type, ClCp'Zr[X-2-R(1)-4-R(2)-6-(Ph(2)P=O)C(6)H(2)](2) (Cp' = C(5)H(5), 2a: X = O, R(1) = Ph, R(2) = H; 2b: X = O, R(1) = F, R(2) = H; 2c: X = O, R(1) = (t)Bu, R(2) = H; 2d: X = O, R(1) = R(2) = (t)Bu; 2e: X = O, R(1) = SiMe(3), R(2) = H; 2f: X = S, R(1) = SiMe(3), R(2) = H; Cp' = C(5)Me(5), 2g: X = O, R(1) = SiMe(3), R(2) = H), have been synthesized in high yields. These complexes were identified by (1)H {(13)C} NMR and elemental analyses. Structures for 2b, 2c and 2f were further confirmed by X-ray crystallography. Structural characterization of these complexes reveals crowded environments around the zirconium. Complexes 2b and 2c adopt six-coordinate, distorted octahedral geometry around the zirconium center, in which the equatorial positions are occupied by three oxygen atoms of two chelating phosphine oxide-bridged phenolate ligands and a chlorine atom. The cyclopentadienyl ring and one oxygen atom of the ligand are coordinated on the axial position. Complex 2f also folds a six-coordinate, distorted octahedral geometry around the Zr center, consisting of a Cp-Zr-O (in P=O) axis [177.16°] and a distorted plane of two sulfur atoms and one oxygen atom of two chelating phosphine oxide-bridged thiophenolate ligands as well as a chlorine atom. When activated by modified methylaluminoxane (MMAO), all the complexes exhibited high activities towards ethylene polymerization at high temperature (75 °C), giving high molecular weight polymers with unimodal molecular weight distribution. The formation of 14-electron, cationic metal alkyl species might come from the Zr-O (in phenol ring) bond cleavage based on the DFT calculations study.

Identification and characterization of a new Zirconium hydride

International Nuclear Information System (INIS)

Zhao, Z.

2007-01-01

In order to control the integrity of the fuel clad, alloy of zirconium, it is necessary to predict the behavior of zirconium hydrides in the environment (temperature, stress...), at a microscopic scale. A characterization study by TEM of hydrides has been realized. It shows little hydrides about 500 nm, in hydride Zircaloy 4. Then a more detailed study identified a new hydride phase presented in this paper. (A.L.B.)

Sorption of cesium on titanium and zirconium phosphates

International Nuclear Information System (INIS)

Lebedev, V.N.; Mel'nik, N.A.; Rudenko, A.V.

2003-01-01

Titanium and zirconium phosphates were prepared from mineral raw materials of the Kola Peninsula. Their capability to recover cesium cations from the model solutions and liquid radioactive waste (LRW) was studied. Titanium phosphate prepared from solutions formed by titanite breakdown demonstrates greater distribution coefficients of cesium as compared to zirconium phosphate. Titanium phosphate as a cheaper agent featuring greater sorption capacity was recommended for treatment of LRW to remove cesium [ru

[The clinical application of zirconium-dioxide-ceramics. Case report].

Science.gov (United States)

Somfai, Dóra; Zsigmond, Ágnes; Károlyházy, Katalin; Kispély, Barbara; Hermann, Péter

2015-12-01

Due to its outstanding physical, mechanical and esthetic properties, zirconium-dioxide is one of the most popular non-metal denture, capable of surpassing PFM in most cases. The recent advances of CAD/CAM technology makes it a good alternitve. Here we show the usefulness of zirconium-dioxide in everyday dental practice through three case reports.

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Hindle, E. D.

1984-01-01

An array of rods is assembled to form a fuel element for a pressurized water reactor, the rods comprising zirconium alloy sheathed nuclear fuel pellets and containing helium. The helium gas pressure is selected for each rod so that it differs substantially from the helium gas pressure in its closest neighbors. In a preferred arrangement the rods are arranged in a square lattice and the helium gas pressure alternates between a relatively high value and a relatively low value so that each rod has as its closest neighbors up to four rods containing helium gas at the other pressure value

Nuclear reactor fuel elements

Energy Technology Data Exchange (ETDEWEB)

Hindle, E. D.

1984-10-16

An array of rods is assembled to form a fuel element for a pressurized water reactor, the rods comprising zirconium alloy sheathed nuclear fuel pellets and containing helium. The helium gas pressure is selected for each rod so that it differs substantially from the helium gas pressure in its closest neighbors. In a preferred arrangement the rods are arranged in a square lattice and the helium gas pressure alternates between a relatively high value and a relatively low value so that each rod has as its closest neighbors up to four rods containing helium gas at the other pressure value.

Peculiarities of formation of zirconium aluminides in hydride cycle mode

International Nuclear Information System (INIS)

Muradyan, G.N.

2016-01-01

The zirconium aluminides are promising structural materials in aerospace, mechanical engineering, chemical industry, etc. They are promising for manufacturing of heat-resistant wires, that will improve the reliability and efficiency of electrical networks. In the present work, the results of study of zirconium aluminides formation in the Hydride Cycle (HC) mode, developed in the Laboratory of high-temperature synthesis of the Institute of Chemical Physics of NAS RA, are described. The formation of zirconium aluminides in HC proceeded according to the reaction xZrH_2+(1-x)Al → alloy Zr_xAl(1-x)+H_2↑. The samples were certified using: chemical analysis to determine the content of hydrogen (pyrolysis method); differential thermal analysis (DTA, derivatograph Q-1500, T_heating = 1000°C, rate 20°C/min); X-ray analysis (XRD, diffractometer DRON-0.5). The influences of the ratio of powders ZrH_2/Al in the reaction mixture, compacting pressure, temperature and heating velocity on the characteristics of the synthesized aluminides were determined. In HC, the solid solutions of Al in Zr, single phase ZrAl_2 and ZrAl_3 aluminides and Zr_3AlH_4.49 hydride were synthesized. Formation of aluminides in HC mode took place by the solid-phase mechanism, without melting of aluminum. During processing, the heating of the initial charge up to 540°C resulted in the decomposition of zirconium hydride (ZrH_2) to HCC ZrH_1.5, that interacted with aluminum at 630°C forming FCC alumohydride of zirconium. Further increase of the temperature up to 800°C led to complete decomposition of the formed alumohydride of zirconium. The final formation of the zirconium aluminide occurred at 1000-1100°C in the end of HC process. Conclusion: in the synthesis of zirconium aluminides, the HC mode has several significant advantages over the conventional modes: lower operating temperatures (1000°C instead of 1800°C); shorter duration (1.5-2 hours instead of tens of hours); the availability of

Distribution of zirconium in the nitric acid-water-TPB-diluent system

International Nuclear Information System (INIS)

Shu, J.; Floh de Araujo, B.

1984-10-01

This paper deals with the extraction behaviour of zirconium in TBP/diluent-HNO 3 -H 2 O systems. The main purpose is to increase the uranium decontamination factor by adjusting the extraction conditions so that zirconium extraction is kept at a mininum. Equilibrium diagram, TBP concentration, aqueous: organic phases ratio, salting-out effects and uranium loading in the organic phase were the main factors studied. All the experiments have been carried out with zirconium in the 10 -2 - 10 -3 M concentration range. The extractant degradation products influence upon ziconium behaviour was also verified. With the data obtained it was possible to introduce some modification in the standard Purex flow-sheet with the increase of the decontamination of uranium from zirconium. 5 refs., 9 figs

Comparison of Zirconium Phosphonate-Modified Surfaces for Immobilizing Phosphopeptides and Phosphate-Tagged Proteins.

Science.gov (United States)

Forato, Florian; Liu, Hao; Benoit, Roland; Fayon, Franck; Charlier, Cathy; Fateh, Amina; Defontaine, Alain; Tellier, Charles; Talham, Daniel R; Queffélec, Clémence; Bujoli, Bruno

2016-06-07

Different routes for preparing zirconium phosphonate-modified surfaces for immobilizing biomolecular probes are compared. Two chemical-modification approaches were explored to form self-assembled monolayers on commercially available primary amine-functionalized slides, and the resulting surfaces were compared to well-characterized zirconium phosphonate monolayer-modified supports prepared using Langmuir-Blodgett methods. When using POCl3 as the amine phosphorylating agent followed by treatment with zirconyl chloride, the result was not a zirconium-phosphonate monolayer, as commonly assumed in the literature, but rather the process gives adsorbed zirconium oxide/hydroxide species and to a lower extent adsorbed zirconium phosphate and/or phosphonate. Reactions giving rise to these products were modeled in homogeneous-phase studies. Nevertheless, each of the three modified surfaces effectively immobilized phosphopeptides and phosphopeptide tags fused to an affinity protein. Unexpectedly, the zirconium oxide/hydroxide modified surface, formed by treating the amine-coated slides with POCl3/Zr(4+), afforded better immobilization of the peptides and proteins and efficient capture of their targets.

The neutron and gamma-ray dose characterization using the Monte Carlo method to study the feasibility of the Prompt Gamma Activation Analysis technique at IPR-R1 TRIGA reactor in Brazil

Energy Technology Data Exchange (ETDEWEB)

Guerra, Bruno T.; Soares, Alexandre L.; Grynberg, Suely E.; Menezes, Maria Angela B.C., E-mail: brunoteixeiraguerra@yahoo.com.br, E-mail: menezes@cdtn.br, E-mail: asleal@cdtn.br, E-mail: seg@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2013-07-01

The IPR-R1 is a reactor type TRIGA, Mark-I model, manufactured by the General Atomic Company and installed at Nuclear Technology Development Centre (CDTN) of Brazilian Nuclear Energy Commission (CNEN), in Belo Horizonte, Brazil. It is a light water moderated and cooled, graphite-reflected, open-pool type research reactor. IPR-R1 works at 100 kW but it will be briefly licensed to operate at 250 kW. It presents low power, low pressure, for application in research, training and radioisotopes production. The fuel is an alloy of zirconium hydride and uranium enriched at 20% in {sup 235}U. The Implementation of the PGNAA (Prompt Gamma Neutron Activation Analysis) Technical at the TRIGA IPR-R1 research reactor of the CDTN will significantly increase in the types of matrices analyzable. A project is underway in order to implement this technique in CDTN. In order of verified the feasibility of the PGNAA at the TRIGA reactor, the MCNP (Monte Carlo N-Particle) method is used to theoretical calculations. This paper presents the results of a preliminary study of the neutron and gamma-ray dose in the room where the reactor is located, in case of implementation of this technique in the IPR-R1. (author)

The neutron and gamma-ray dose characterization using the Monte Carlo method to study the feasibility of the Prompt Gamma Activation Analysis technique at IPR-R1 TRIGA reactor in Brazil

International Nuclear Information System (INIS)

Guerra, Bruno T.; Soares, Alexandre L.; Grynberg, Suely E.; Menezes, Maria Angela B.C.

2013-01-01

The IPR-R1 is a reactor type TRIGA, Mark-I model, manufactured by the General Atomic Company and installed at Nuclear Technology Development Centre (CDTN) of Brazilian Nuclear Energy Commission (CNEN), in Belo Horizonte, Brazil. It is a light water moderated and cooled, graphite-reflected, open-pool type research reactor. IPR-R1 works at 100 kW but it will be briefly licensed to operate at 250 kW. It presents low power, low pressure, for application in research, training and radioisotopes production. The fuel is an alloy of zirconium hydride and uranium enriched at 20% in 235 U. The Implementation of the PGNAA (Prompt Gamma Neutron Activation Analysis) Technical at the TRIGA IPR-R1 research reactor of the CDTN will significantly increase in the types of matrices analyzable. A project is underway in order to implement this technique in CDTN. In order of verified the feasibility of the PGNAA at the TRIGA reactor, the MCNP (Monte Carlo N-Particle) method is used to theoretical calculations. This paper presents the results of a preliminary study of the neutron and gamma-ray dose in the room where the reactor is located, in case of implementation of this technique in the IPR-R1. (author)

Experimental studies of U-Pu-Zr fast reactor fuel pins in the Experimental Breeder Reactor 2

International Nuclear Information System (INIS)

Pahl, R.G.; Porter, D.L.; Lahm, C.E.; Hofman, G.L.

1990-01-01

Argonne National Laboratory's Integral Fast Reactor (IFR) concept has been under demonstration in the Experimental Breeder Reactor II (EBR-II) since February 1985. Irradiation tests of U-Zr and U-Pu-Zr fuel pins to >15 at. pct burnup have demonstrated their viability as driver fuel prototypes in innovative design liquid metal reactors. A number of technically challenging irradiation effects have been observed and are now under study. Microstructural changes in the fuel are dominated early in exposure by grain boundary cavitation and fission gas bubble growth, producing large amounts of swelling. Irradiation creep and swelling of the austenitic (D9) and martensitic (HT-9) candidate cladding alloys have been measured and correlate well with property modeling efforts. Chemical interaction between the fuel and cladding alloys has been characterized to assess the magnitude of cladding wastage during steady-state irradiation. Significant interdiffusion of the uranium and zirconium occurs producing metallurgically distinct zones in the fuel

Scram device for atomic reactors

International Nuclear Information System (INIS)

Noyes, R.C.; Zaman, S.O.; Stuteville, D.W.

1978-01-01

A sensor chamber having one cavity containing coolant separated by a diaphragm from another cavity containing a fixed mass of inert gas is located within a safety assembly of a liquid metal-cooled nuclear reactor. The liquid cavity is in fluid communication with the coolant outside the chamber through a flow limiting orifice. An actuating bellows in fluid communication with the gas cavity is in contact with coolant outside the chamber and is connected to a push rod, which serves as a trigger for a poison bundle release mechanism. During slow changes in reactor coolant pressure experienced under normal operation, the diaphragm moves to equalize the gas cavity and liquid cavity pressures with the coolant pressure outside the chamber. The actuating bellows does not move because it is biased so that a threshold pressure difference is required before it will expand. Under a more rapid drop in coolant pressure, such as is associated with a loss of forced flow, the threshold is overcome and the actuating bellows will also move, thereby triggering the release mechanism to shut down the reactor. In an alternate embodiment, the actuating bellows is connected to the liquid cavity rather than to the gas cavity. (Auth.)

Separation process of zirconium and hafnium

International Nuclear Information System (INIS)

Hure, J.; Saint-James, R.

1955-01-01

About the separation different processes of zirconium-hafnium, the extraction by solvent in cross-current is the most easily the process usable on an industrial scale. It uses tributyl phosphate as solvent, diluted with white spirit to facilitate the decanting. Some exploratory tests showed that nitric environment seemed the most favorable for extraction; but a lot of other factors intervene in the separation process. We studied the influence of the acidity successively, the NO 3 - ions concentration, the role of the cation coming with NO 3 - , as well as the influence of the concentration of zirconium in the solution on the separation coefficient β = α Zr / α Hf . (M.B.) [fr

Zirconium diselenite microstructures, formation and mechanism

Science.gov (United States)

Naik, Chandan C.; Salker, A. V.

2018-04-01

In this work, a series of microstructures of zirconium diselenite (Zr(SeO3)2) has been prepared via a simple precipitation method at room temperature without adding any organic surfactants. Phase purity of the sample has been checked by X-ray Diffraction. From the SEM, FESEM, and TEM images spheroid nanoparticles to the starfish-like structure of zirconium diselenite are detected. The morphological evolution processes were investigated carefully following time-dependent experiments and a growth mechanism has been proposed. Two different crystal growth processes, the oriented attachment process accompanying the Ostwald ripening process were held responsible for the formation of a structure resembling starfish having four arms.

Design of an organic simplified nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Shirvan, Koroush [Dept. of Nuclear Science and Engineering, Massachusetts Institute of Technology, Cambridge (United States); Forrest, Eric [Primary Standards Laboratory, Sandia National Laboratories, Albuquerque (United States)

2016-08-15

Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

Design of an Organic Simplified Nuclear Reactor

Directory of Open Access Journals (Sweden)

Koroush Shirvan

2016-08-01

Full Text Available Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

Study of the processes for of remelting zirconium alloys in an electric arc furnace

Energy Technology Data Exchange (ETDEWEB)

Pereira, Luiz A.T.; Rossi, Jesualdo L.; Costa, Guilherme R.; Martinez, Luis G.; Sato, Ivone M., E-mail: luiz.atp@uol.com.br, E-mail: jelrossi@ipen.br, E-mail: guilhermeramoscosta@gmail.com, E-mail: lgallego@ipen.br, E-mail: imsato@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2015-07-01

Zirconium alloy tubes are used as cladding for fuel elements of PWR nuclear reactors, which contains the UO{sub 2} pellets. In the manufacture of these fuel element parts, machining chips from the nuclear grade zirconium alloys are generated. Hence, these machining chips cannot be discarded, as ordinary metallic waste. Thus, the recycling of this material is a strategic aspect for the nuclear technology, both for economic and environmental issues. The main reason is that nuclear grade alloys have very high cost, are not commercially produced in Brazil and has to be imported for the manufacture of the nuclear fuels. This work discusses a method to melt and recycle Zircaloy chips, using an electric-arc furnace to obtain small laboratory ingots. The chemical composition of the ingots was determined using X-ray fluorescence spectroscopy and was compared to the specifications of nuclear grade Zircaloy and to the chemical composition of the received machining chips. The ingots were annealed in high vacuum, as well as were hot rolled in a mill. The microstructures were characterized by optical microscopy. The hardness was evaluated using the Rockwell B scale hardness. The results showed that the compositions of the recycled Zircaloy comply with the chemical specifications and a suitable microstructure has been obtained for nuclear use. (author)

Susceptibility of cold-worked zirconium-2.5 wt% niobium alloy to delayed hydrogen cracking

International Nuclear Information System (INIS)

Coleman, C.E.

1976-01-01

Notched tensile specimens of cold-worked zirconium-2.5 wt% niobium alloy have been stressed at 350 K and 520 K. At 350 K, above a possible threshold stress of 200 MPa, specimens exhibited delayed failure which was attributed to hydride cracking. Metallography showed that hydrides accumulated at notches and tips of growing cracks. The time to failure appeared to be independent of hydrogen content over the range 7 to 100 ppm hydrogen. Crack growth rates of about 10 -10 m/s deduced from fractography were in the same range as those necessary to fracture pressure tubes. The asymptotic stress intensity for delayed failure, Ksub(1H), appeared to be about 5 MPa√m. With this low value of Ksub(1H) small surface flaws may propagate in pressure tubes which contain large residual stresses. Stress relieving and modified rolling procedures will reduce the residual stresses to such an extent that only flaws 12% of the wall thickness or greater will grow. At 520 K no failures were observed at times a factor of three greater than times to failure at 350 K. Zirconium-2.5 wt% niobium appears to be safe from delayed hydrogen cracking at the reactor operating temperature. (author)

Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

Science.gov (United States)

Clement, J. D.; Rust, J. H.

1977-01-01

Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

Hydrogen content in titanium and a titanium–zirconium alloy after acid etching

Energy Technology Data Exchange (ETDEWEB)

Frank, Matthias J.; Walter, Martin S. [Department of Biomaterials, Institute for Clinical Dentistry, University of Oslo, P.O. Box 1109, Blindern, NO-0317 Oslo (Norway); Institute of Medical and Polymer Engineering, Chair of Medical Engineering, Technische Universität München, Boltzmannstrasse 15, 85748 Garching (Germany); Lyngstadaas, S. Petter [Department of Biomaterials, Institute for Clinical Dentistry, University of Oslo, P.O. Box 1109, Blindern, NO-0317 Oslo (Norway); Wintermantel, Erich [Institute of Medical and Polymer Engineering, Chair of Medical Engineering, Technische Universität München, Boltzmannstrasse 15, 85748 Garching (Germany); Haugen, Håvard J., E-mail: h.j.haugen@odont.uio.no [Department of Biomaterials, Institute for Clinical Dentistry, University of Oslo, P.O. Box 1109, Blindern, NO-0317 Oslo (Norway)

2013-04-01

Dental implant alloys made from titanium and zirconium are known for their high mechanical strength, fracture toughness and corrosion resistance in comparison with commercially pure titanium. The aim of the study was to investigate possible differences in the surface chemistry and/or surface topography of titanium and titanium–zirconium surfaces after sand blasting and acid etching. The two surfaces were compared by X-ray photoelectron spectroscopy, secondary ion mass spectroscopy, scanning electron microscopy and profilometry. The 1.9 times greater surface hydrogen concentration of titanium zirconium compared to titanium was found to be the major difference between the two materials. Zirconium appeared to enhance hydride formation on titanium alloys when etched in acid. Surface topography revealed significant differences on the micro and nanoscale. Surface roughness was increased significantly (p < 0.01) on the titanium–zirconium alloy. High-resolution images showed nanostructures only present on titanium zirconium. - Highlights: ► TiZr alloy showed increased hydrogen levels over Ti. ► The alloying element Zr appeared to catalyze hydrogen absorption in Ti. ► Surface roughness was significantly increased for the TiZr alloy over Ti. ► TiZr alloy revealed nanostructures not observed for Ti.