DOCSLIB.ORG

Part Ii: Technical Analysis and Systems Study

Total Page：16

File Type：pdf, Size：1020Kb

Download full-text PDF Read full-text
Part Ii: Technical Analysis and Systems Study PART II: TECHNICAL ANALYSIS AND SYSTEMS STUDY 111 1. PARTITIONING 1.1 Aqueous separation techniques This section briefly describes aqueous separation techniques currently used on industrial scale and research activities in the field of new separation methods for more effective separation of minor actinides and fission products. There has been a large number of reports published until now and a selection of the important ones is listed in Annex D. 1.1.1 PUREX process The PUREX process, see Figure II.1, which is universally employed in the irradiated fuel reprocessing industry, is a wet chemical process based on the use of TBP, a solvent containing phosphorus. As shown in Table II.1 this solvent displays the property of extracting actinide cations in even oxidation states IV and VI, in the form of a neutral complex of the type M•An•2TBP (where M is the metallic cation and A an anion, generally nitrate ion), from an acidic aqueous medium. Conversely, the actinide cations with odd oxidation state are not significantly extracted, at least in the high acidity conditions prevailing during reprocessing operations. Uranium and plutonium, whose stable oxidation states in nitric medium are VI and IV, respectively, are co-extracted by TBP and thus separated from the bulk of the fission products which remain in the aqueous phase. This is the basic principle of the PUREX process. Table II.1 Extractability of actinide nitrates in 3 M nitric acid by TBP Oxidation state III IV V VI U (m) (l) m Np (m) l m Pu (l) m (l) (m) Am l (m) Cm l m: extractable by TBP, l: not extractable by TBP, ( ): unstable in the media Uranium and plutonium are recovered with an industrial yield close to 99.9% (including losses in secondary wastes). 113 1.1.1.1 Minor actinides Americium and curium Among the minor actinides, americium and curium, which are stable in valency III, are not extracted by TBP and remain in the aqueous phase. They accordingly follow the path of the fission products and are currently managed like the latter by conditioning in a glass matrix. Neptunium Another minor actinide, neptunium, whose stable oxidation state is V, is hence very slightly extractable in this species by TBP. However, in the chemical conditions of the first cycle extraction operation of the PUREX process (presence of nitrous acid), part of the Np(V) is oxidised to VI, and accordingly extracted in the organic phase. The operating results of the UP3 plant reveal that the majority of the Np is extracted by TBP, follows the uranium stream, and is separated from the latter in the second uranium purification cycle. The effluent containing neptunium is currently added to the high-level waste stream. Hence all the neptunium is sent to vitrification apart from the proportion following the Pu product. The behaviour of neptunium is independent of the type of fuel reprocessed. 1.1.1.2 Long-lived fission products Among the fission products with long-lived isotopes, three elements (technetium, zirconium and iodine) display specific behaviour in the PUREX process, which could be exploited for their separation (see Figure II.1). Technetium During fuel dissolution, part of the technetium, probably in metallic or oxide form, does not go into solution. This fraction, estimated at 10 to 20% of the total Tc for a UO2 fuel, accompanies the “insoluble residues” essentially consisting of noble metals (Ru, Rh, Pd). These insoluble residues are currently incorporated in the vitrified or cemented wastes. In nitric medium and in the absence of a reducing agent, the dissolved technetium is in its highest valency (VII) which is the most stable. It occurs in anionic form in the state of pertechnetate ion, - TcO 4 . This species can be extracted by TBP at the same time as a metallic cation by substitution of a pertechnetate group for a nitrate group in the neutral complex extracted. The occurrence of this co-extraction was observed with the main metallic cations extracted by TBP, and particularly with the cation ZrO2+ (see below), which is present in large amounts in the feed solution to the extraction cycle. This extraction of Tc proved to be a serious hindrance because of its interference with the chemical mechanisms of the partitioning operation (U/Pu separation). The PUREX process accordingly had to be adapted to limit the extraction of Tc. 114 Figure II.1 Behaviour of long-lived elements in PUREX process (An example of UP3 La Hague) Hulls and end pieces Activation I (>90%) products Dissolution Off-gas treatment U/Pu Purification 1st Cycle FP Tc U/Pu Extraction U scrubbing scrubbing separation stripping Am,Cm Np (minor fraction) Insolubles Cs, Zr (100%) Tc (up to 90%) U Np (major fraction) U purification Pu Vitrification Pu purification Tc (>10%) The final flow chart, which includes a specific stripping step for the Tc extracted by the TBP, produces an effluent containing a large fraction of the element. This effluent is currently added to the main fission product stream. Zirconium Zirconium, which is present in the form of the ion ZrO2+, can be extracted by TBP, but to a lesser degree than uranium and plutonium. This element is effectively stripped in the “FP washing” operation, which immediately follows extraction. The aqueous phase from the washing must be recycled to extraction due to the large amounts of U and Pu which it contains. In the present status of the process, zirconium is not specifically isolated. Iodine Iodine presents a special case in so far as this element is extremely volatile in the elemental state. This property is exploited for the containment of this fission product in the “process head end”, 115 thus preventing its dispersion in the downstream operations where its controlled management would be quite difficult. The operating conditions of fuel dissolution are selected to ensure that the iodine is brought to and maintained in the elemental state, and to entrain it in the off-gas. The iodine is recovered in an aqueous solution by caustic scrub of the off-gas. This specific effluent of iodine, which is discharged into the sea today, thus contains nearly all the iodine initially present in the irradiated fuel. Some reprocessing plants envisage the use of iodine immobilisation by adsorption on silver impregnated zeolites. Other long-lived fission products As to the other long-lived fission products, it is clear that the PUREX process cannot be used to separate caesium and strontium, since these mono- and divalent elements are unextractable by TBP. The behaviour in the PUREX process of the other fission products which have long-lived isotopes (Pd, Se, Sn) is not precisely known. A combined electrolytic extraction of Pd2+ with the other platinum group 3+ 3+ - 2- elements (RuNO , Rh ) and TcO 4 (and probably SeO 4 ) seems to be promising from even higher acidic PUREX liquors [1]. 1.1.1.3 Long-lived activation products The activation products formed in the fuel element structural metals (stainless steels, inconel and zircaloy) mostly remain in these materials and are found in the corresponding “hulls and end pieces” waste stream. The 14C issue should receive increasing attention because of this isotope’s impact on the biosphere. 1.1.1.4 Conclusions The behaviour of the minor actinides and long-lived fission products in the PUREX process can be divided into three categories: · elements already partially separated by the PUREX process: neptunium, technetium and iodine. For these elements, the R&D objective involves process extensions to achieve the desired separation performance. This first aspect is discussed further in Section 1.1.2. · elements separable by TBP, for which a complementary step to the present PUREX process can be developed. This applies to zirconium. · elements that cannot be separated by the PUREX process: – americium and curium, – caesium, strontium and probably the other fission products (Pd, Se, Sn). To separate these elements, it is necessary to develop new classes of extractants, or to resort to different separation methods. The corresponding developments are discussed in Section 1.1.3 (Am and Cm) and Section 1.1.4 (FPs). 116 1.1.2 Improved separation of long-lived elements in PUREX process 1.1.2.1 Neptunium Neptunium is present in the irradiated fuel dissolution liquor (nitric acid medium) in oxidation states V and VI. The extractability of Np(V) by tributylphosphate (TBP) organic solution (the PUREX solvent) is rather poor whereas that of Np(VI) is good, approaching that of U(VI) and Pu(IV). Hence one alternative to separate neptunium from the wastes is to extract this element in the first cycle of the PUREX process together with U(VI) and Pu(IV). This requires the oxidation of the whole neptunium inventory to the VI oxidation state, allowing Np extraction by TBP. Thus the redox reaction to convert Np(V) to Np(VI) is the key point to be addressed to achieve ~100% Np extraction in the first cycle. Np(V) can be oxidised to Np(VI) by various means including: · chemical oxidation using nitric/nitrous acid mixture, like those present in the fuel dissolution liquors, intentionally adding oxidants such as vanadium (V) compounds; · using force field oxidation like: b, g radiolysis oxidation, photochemical oxidation, sonochemical oxidation. After its co-extraction with U and Pu, neptunium can be selectively separated from these elements, either using the regular PUREX cycles (for example in the second uranium cycle) or under the action of specific reagents like butyraldehydes. Computer codes of the PUREX process are available for calculating the behaviour of Np in the PUREX extraction cycles. The current research trend in this area is to define more refined chemical models for Np behaviour that are more elaborate than those hitherto employed.
Load more

Recommended publications
  • Spent Fuel Reprocessing
    Spent Fuel Reprocessing Robert Jubin Oak Ridge National Laboratory Reprocessing of used nuclear fuel is undertaken for several reasons. These include (1) recovery of the valuable fissile constituents (primarily 235U and plutonium) for subsequent reuse in recycle fuel; (2) reduction in the volume of high-level waste (HLW) that must be placed in a geologic repository; and (3) recovery of special isotopes. There are two broad approaches to reprocessing: aqueous and electrochemical. This portion of the course will only address the aqueous methods. Aqueous reprocessing involves the application of mechanical and chemical processing steps to separate, recover, purify, and convert the constituents in the used fuel for subsequent use or disposal. Other major support systems include chemical recycle and waste handling (solid, HLW, low-level liquid waste (LLLW), and gaseous waste). The primary steps are shown in Figure 1. Figure 1. Aqueous Reprocessing Block Diagram. Head-End Processes Mechanical Preparations The head end of a reprocessing plant is mechanically intensive. Fuel assemblies weighing ~0.5 MT must be moved from a storage facility, may undergo some degree of disassembly, and then be sheared or chopped and/or de-clad. The typical head-end process is shown in Figure 2. In the case of light water reactor (LWR) fuel assemblies, the end sections are removed and disposed of as waste. The fuel bundle containing the individual fuel pins can be further disassembled or sheared whole into segments that are suitable for subsequent processing. During shearing, some fraction of the radioactive gases and non- radioactive decay product gases will be released into the off-gas systems, which are designed to recover these and other emissions to meet regulatory release limits.
    [Show full text]
  • Development of a Solvent Extraction Process for Group Actinide Recovery from Used Nuclear Fuel
    THESIS FOR THE DEGREE OF DOCTOR OF PHILOSOPHY Development of a Solvent Extraction Process for Group Actinide Recovery from Used Nuclear Fuel EMMA H. K. ANEHEIM Department of Chemical and Biological Engineering CHALMERS UNIVERSITY OF TECHNOLOGY Gothenburg, Sweden, 2012 Development of a Solvent Extraction Process for Group Actinide Recovery from Used Nuclear Fuel EMMA H. K. ANEHEIM ISBN 978-91-7385-751-2 © EMMA H. K. ANEHEIM, 2012. Doktorsavhandlingar vid Chalmers tekniska högskola Ny serie Nr 3432 ISSN 0346-718X Department of Chemical and Biological Engineering Chalmers University of Technology SE-412 96 Gothenburg Sweden Telephone + 46 (0)31-772 1000 Cover: Radiotoxicity as a function of time for the once through fuel cycle (left) compared to one P&T cycle using the GANEX process (right) (efficiencies: partitioning from Table 5.5.4, transmutation: 99.9%). Calculations performed using RadTox [HOL12]. Chalmers Reproservice Gothenburg, Sweden 2012 Development of a Solvent Extraction Process for Group Actinide Recovery from Used Nuclear Fuel EMMA H. K. ANEHEIM Department of Chemical and Biological Engineering Chalmers University of Technology Abstract When uranium is used as fuel in nuclear reactors it both undergoes neutron induced fission as well as neutron capture. Through successive neutron capture and beta decay transuranic elements such as neptunium, plutonium, americium and curium are produced in substantial amounts. These radioactive elements are mostly long-lived and contribute to a large portion of the long term radiotoxicity of the used nuclear fuel. This radiotoxicity is what makes it necessary to isolate the used fuel for more than 100,000 years in a final repository in order to avoid harm to the biosphere.
    [Show full text]
  • Reactions of Lithium Nitride with Some Unsaturated Organic Compounds. Perry S
    Louisiana State University LSU Digital Commons LSU Historical Dissertations and Theses Graduate School 1963 Reactions of Lithium Nitride With Some Unsaturated Organic Compounds. Perry S. Mason Jr Louisiana State University and Agricultural & Mechanical College Follow this and additional works at: https://digitalcommons.lsu.edu/gradschool_disstheses Recommended Citation Mason, Perry S. Jr, "Reactions of Lithium Nitride With Some Unsaturated Organic Compounds." (1963). LSU Historical Dissertations and Theses. 898. https://digitalcommons.lsu.edu/gradschool_disstheses/898 This Dissertation is brought to you for free and open access by the Graduate School at LSU Digital Commons. It has been accepted for inclusion in LSU Historical Dissertations and Theses by an authorized administrator of LSU Digital Commons. For more information, please contact [email protected]. This dissertation has been 64—5058 microfilmed exactly as received MASON, Jr., Perry S., 1938- REACTIONS OF LITHIUM NITRIDE WITH SOME UNSATURATED ORGANIC COMPOUNDS. Louisiana State University, Ph.D., 1963 Chemistry, organic University Microfilms, Inc., Ann Arbor, Michigan Reproduced with permission of the copyright owner. Further reproduction prohibited without permission. Reproduced with permission of the copyright owner. Further reproduction prohibited without permission. Reproduced with permission of the copyright owner. Further reproduction prohibited without permission. REACTIONS OF LITHIUM NITRIDE WITH SOME UNSATURATED ORGANIC COMPOUNDS A Dissertation Submitted to the Graduate Faculty of the Louisiana State University and Agricultural and Mechanical College in partial fulfillment of the requireiaents for the degree of Doctor of Philosophy in The Department of Chemistry by Perry S. Mason, Jr. B. S., Harding College, 1959 August, 1963 Reproduced with permission of the copyright owner. Further reproduction prohibited without permission.
    [Show full text]
  • High-Quality, Low-Cost Bulk Gallium Nitride Substrates
    ADVANCED MANUFACTURING OFFICE High-Quality, Low- Cost Bulk Gallium Nitride Substrates Electrochemical Solution Growth: A Scalable Semiconductor Manufacturing Process The ever-growing demand in the past decade for more energy efficient solid-state lighting and electrical power conversion is leading to a higher demand for wide bandgap semiconductor-based devices, such as gallium nitride (GaN), over traditional silicon (Si)-based devices. High cost and limited availability, how- ever, have hindered the adoption of GaN substrates to date. When utilizing GaN, current LED and power electronic device applications employ GaN epitaxially grown on top of non-GaN substrates. The lattice mismatch between the epitaxial GaN layer and the non-native substrate surface leads to con- siderable stress and high defect densities, ultimately compromising device yield and Conceptual diagram of the ESG reactor. Photo courtesy of Sandia National Laboratories performance. While bulk growth of GaN can combat these issues, current growth methods for bulk GaN have not fostered widespread adoption to date due to lim- This project will help develop ESG into a viable GaN bulk growth process that is well ited scalability, low material quality, high suited for scalability to large-area wafer manufacturing. Bulk GaN is important to bol- operating temperatures and pressures, stering U.S. competitiveness in high-efficiency power electronics and solid-state lighting. and slow growth rates. A fundamentally different manufacturing route for bulk Benefits for Our Industry and Our Nation growth of GaN not driven by thermal pro- Scaling the ESG growth method to large area GaN crystals could reduce the production cesses is needed to provide an adequate cost of bulk GaN wafers by up to a factor of 10.
    [Show full text]
  • Engineering Design of Aries-Iii*
    PAPER ENGINEERING DESIGN OF ARIES-HI* O.K. Sze, Argonne National Laboratory, Argonne, IL 60439 C. Wong, General Atomics, San Diego, CA 92138 E. Cheng, TSI Research, Solana Beach, CA 92075 M.E. Sawan, I.N. Sviatoslavsky, J.P. Blanchard, L.A. El-Guebaly, H.Y. Khater, E.A. Mogahed University of Wisconsin, Madison, Wl 53706 P. Gierszewski, Canadian Fusion Fuels Tech. Prog., Mississauga, Ontario, Canada, L5J1K3 R. Hollies, Hollies Fluids Consulting, Pinawa, MB, ROE 1 L0 Canada and the ARIES Team M Z. <*• « (5 E B RECEIVE. AUG 2 6 1993 o M c -5 «3 OSTI Th« uibmitted manuscript has been authored by a contractor of the U. 5. Government under contract No. W-31-109-ENG-38- l Accordingly, the U. S. Government retains a nonexclusive, royalty-free license to publish or reproduce tha published form of this contribution, or allow others to do so, far U. 5. Government purposes. July 1993 •so * Work supported by the U.S. Department of Energy, Office of Fusion Energy, under Contract No. W-31-109-Eng-38. Paper to be presented at the Second Wisconsin Symposium on 3He and Fusion Power, University of Wisconsin, Madison, Wl, July 19-21,1993. DWTFMBUTION OF THIS DOOUMENT IS UNLIMITED ENGINEERING DESIGN OF ARIES-III* D.K. Sze, Argonne National Laboratory, Argonne, IL 60439 C. Wong, General Atomics, San Diego, CA 92138 E. Cheng, TSI Research, Solana Beach, CA 92075 M.E. Sawan, I.N. Sviatoslavsky, J.P. Blanchard, LA. El-Guebaly, H.Y. Khater, E.A. Mogahed University of Wisconsin, Madison, Wl 53706 P.
    [Show full text]
  • Naming Compounds Practice Problems KEY Naming Simple Ionic Compounds
    Chemistry HS/Science Unit: 05 Lesson: 01 Naming Compounds Practice Problems KEY Naming Simple Ionic Compounds Name the following compounds: 1) KCl potassium chloride 2) MgI2 magnesium iodide 3) FeO iron (II) oxide 4) Fe2O3 iron (III) oxide 5) Cu3P copper (I) phosphide 6) SnSe2 tin (IV) selenide 7) TiBr3 titanium (III) bromide 8) GaAs gallium arsenide 9) BeF2 beryllium fluoride 10) Cs3N cesium nitride Write the formulas for the following compounds: 1) lithium iodide LiI 2) cobalt (III) oxide Co2O3 3) calcium fluoride CaF2 4) silver bromide AgBr 5) sodium hydride NaH 6) vanadium (V) sulfide V2S5 7) lead (II) nitride Pb3N2 8) titanium (II) selenide TiSe 9) manganese (VII) arsenide Mn3As7 10) gallium chloride GaCl3 ©2013, TESCCC 06/17/13 page 1 of 4 Chemistry HS/Science Unit: 05 Lesson: 01 Naming Compounds Practice Problems Naming Complex (polyatomic) Ionic Compounds Name the following compounds: 1) NH4Cl ammonium chloride 2) Fe(NO3)3 iron (III) nitrate 3) Pb(SO4)2 lead (IV) sulfate 4) Ag3PO4 silver phosphate 5) Be(HCO3)2 beryllium hydrogen carbonate 6) Al(CN)3 aluminum cyanide 7) Mn2(SO3)3 manganese (III) sulfite 8) Sr(C2H3O2)2 strontium acetate 9) Ti(CN)4 titanium (IV) cyanide 10) YClO3 yttrium chlorate Write the formulas for the following compounds: 1) lead (IV) sulfate Pb(SO4)2 2) silver cyanide AgCN 3) copper (II) chlorate Cu(ClO3)2 4) chromium (IV) phosphate Cr3(PO4)4 5) vanadium (IV) carbonate V(CO3)2 6) ammonium oxide (NH4)2O 7) tin (II) nitrite Sn(NO2)2 8) chromium (III) hydroxide Cr(OH)3 9) titanium (II) acetate Ti(C2H3O2)2 10)
    [Show full text]
  • Dr. Starr of Atomics International New Head of American Nuclear Societ Y Group Includes Members from 29 Countries Devoted to Ad- Vancement of Nuclear Scienc E
    VOL. 1 NO. 2 ATOMICS INTERNATIONAL a division of North American Aviation, Inc . SEPTEMBER 1958 Dr. Starr Of Atomics International New Head Of American Nuclear Societ y Group Includes Members From 29 Countries Devoted To Ad- vancement Of Nuclear Scienc e At the annual meeting in June of the American Nuclear Society (ANS) . Dr. Chauncey Starr, general manager of Atomics International and vice president of North American Aviation, Inc., was elected president of the society to succeed Dr . Leland J . Haworth . Direc- tor of the Brookhaven National Laboratory . The American Nuclear Society was formed in 1955 to advance nuclear science and engineering . to encour- age research . establish scholarships. and to disseminate technical information . Membership of 3,000 Includes 29 Countries The organization's membership of 3 .000 includes about 140 members from 29 countries outside the United States representing many branches of science and technology . Members are associated with educa. tional , governmental, and research institutions, and with government contractors and industrial companies. First Student Award Establishe d In keeping with one of the objectives of the society, the first annual graduate student award has been estab- lished for the most outstanding graduate student work- ing in the nuclear sciences . NEW ANS OFFICERS-At the recent meeting of the URER ; Dr. Chauncey Starr, general manager of Atomics The award is the first of its kind by the society and American Nuclear Societe , new officers elected are : (felt International and vice-president of North American Avia- to right) John A . Swartnut, deputy director of Oak Ridge tion, Inc. PRESIDENT ; and Octave J.
    [Show full text]
  • Boron-Nitride-Datasheet-V1
    A Unique Proposition for Industry The introduction of boron nitride into our stable of 2D materials widens the focus of delivering innovative solutions for industry Neill Ricketts, CEO Our latest 2D success story Hexotene is a few-layer hexagonal boron nitride (h-BN) nanoplatelet powder with large lateral dimensions. With high chemical purity and mono-layer particles confirmed, Hexotene is the latest addition to our high performance 2D product range. It’s unique characteristics, specifically with regards to electrical conductivity, show some markedly different properties when compared to graphene. This is particularly promising for combined projects using both graphene and boron nitride. Example BN Nano platelets Information Property Measurement Method Boron 42 è 2.0At.% Predominantly few-layer with some Raman Nitrogen 45 è 2.0At.% Layers mono-layer and bi-layer spectroscopy Oxygen 3.0 è 1.0At.% Lateral Carbon 8.0 è 2.0At.% up to 5.0 µm SEM dimensions Method XPS IR UV Wavelenght (nm) 4096 1024 256 Band Gap ~6eV Energy BORON NITRIDE GRAPHENE Graphene Metals Semi- Plastics Boron Conductor Nitride 0.125 0.25 0.50 1.00 2.00 4.00 8.00 Conductors Semi-Conductors Insulators Energy (eV) Diagram showing extremes of electrical conductivity, with Hexagonal boron nitride is white in colour and is unusual as boron nitride having an ultra-wide bandgap. it absorbs high energy UV light, whereas graphene absorbs all light frequencies. What is h-BN? Boron and nitrogen are neighbours of carbon in boron nitride (h-BN). It also happens to be the the Periodic Table. Just as carbon can exist as softest of the BN polymorphs.
    [Show full text]
  • Techniques of Preparation and Crystal Chemistry of Transuranic Chalcogenides and Pnictides D
    Techniques of preparation and crystal chemistry of transuranic chalcogenides and pnictides D. Damien, R. Haire, J. Peterson To cite this version: D. Damien, R. Haire, J. Peterson. Techniques of preparation and crystal chemistry of transuranic chalcogenides and pnictides. Journal de Physique Colloques, 1979, 40 (C4), pp.C4-95-C4-100. 10.1051/jphyscol:1979430. jpa-00218826 HAL Id: jpa-00218826 https://hal.archives-ouvertes.fr/jpa-00218826 Submitted on 1 Jan 1979 HAL is a multi-disciplinary open access L’archive ouverte pluridisciplinaire HAL, est archive for the deposit and dissemination of sci- destinée au dépôt et à la diffusion de documents entific research documents, whether they are pub- scientifiques de niveau recherche, publiés ou non, lished or not. The documents may come from émanant des établissements d’enseignement et de teaching and research institutions in France or recherche français ou étrangers, des laboratoires abroad, or from public or private research centers. publics ou privés. JOURNAL DE PHYSIQUE Colloque C4, supplément au n° 4, Tome 40, avril 1979, page C4-95 Techniques of preparation and crystal chemistry of transuranic chalcogenides and pnictides (*) D. A. Damien (**), R. G. Haire and J. R. Peterson (t) Transuranium Research Laboratory, Oak Ridge National Laboratory, Oak Ridge, TN 37830, U.S.A. Résumé. — La cristallochimie d'un certain nombre de chalcogénures et pnictures d'éléments transuraniens a été étudiée en utilisant les isotopes Am, Cm, Bk et Cf. Les composés ont été préparés par réaction directe du métal avec l'élément chalcogène ou pnictogène à température élevée. Les chalcogénures supérieurs ont été dissociés thermiquement pour donner des composés à stœchiométrie plus basse dont les sesquichalco- génures constituent la limite.
    [Show full text]
  • Carbides and Nitrides of Zirconium and Hafnium
    materials Review Carbides and Nitrides of Zirconium and Hafnium Sergey V. Ushakov 1,* , Alexandra Navrotsky 1,* , Qi-Jun Hong 2,* and Axel van de Walle 2,* 1 Peter A. Rock Thermochemistry Laboratory and NEAT ORU, University of California at Davis, Davis, CA 95616, USA 2 School of Engineering, Brown University, Providence, RI 02912, USA * Correspondence: [email protected] (S.V.U.); [email protected] (A.N.); [email protected] (Q.-J.H.); [email protected] (A.v.d.W.) Received: 6 August 2019; Accepted: 22 August 2019; Published: 26 August 2019 Abstract: Among transition metal carbides and nitrides, zirconium, and hafnium compounds are the most stable and have the highest melting temperatures. Here we review published data on phases and phase equilibria in Hf-Zr-C-N-O system, from experiment and ab initio computations with focus on rocksalt Zr and Hf carbides and nitrides, their solid solutions and oxygen solubility limits. The systematic experimental studies on phase equilibria and thermodynamics were performed mainly 40–60 years ago, mostly for binary systems of Zr and Hf with C and N. Since then, synthesis of several oxynitrides was reported in the fluorite-derivative type of structures, of orthorhombic and cubic higher nitrides Zr3N4 and Hf3N4. An ever-increasing stream of data is provided by ab initio computations, and one of the testable predictions is that the rocksalt HfC0.75N0.22 phase would have the highest known melting temperature. Experimental data on melting temperatures of hafnium carbonitrides are absent, but minimum in heat capacity and maximum in hardness were reported for Hf(C,N) solid solutions.
    [Show full text]
  • Organic Liquids As Reactor Codants and Moderators
    <" . • t téLb TECHNICAL REPORTS SERIES No. 70 Organic Liquids as Reactor Codants and Moderators ?J INTERNATIONAL ATOMIC ENERGY AGENCY, VIENNA, 1967 ORGANIC LIQUIDS AS REACTOR COOLANTS AND MODERATORS The following States are Members of the International Atomic Energy Agency: AFGHANISTAN GABON NICARAGUA ALBANIA GERMANY, FEDERAL NIGERIA ALGERIA REPUBLIC OF NORWAY ARGENTINA GHANA PAKISTAN AUSTRALIA GREECE PANAMA AUSTRIA GUATEMALA PARAGUAY BELGIUM HAITI PERU BOLIVIA HOLY SEE PHILIPPINES BRAZIL HONDURAS POLAND BULGARIA HUNGARY PORTUGAL BURMA ICELAND ROMANIA BYELORUSSIAN SOVIET INDIA SAUDI ARABIA SOCIALIST REPUBLIC INDONESIA SENEGAL CAMBODIA IRAN SOUTH AFRICA CAMEROON IRAQ SPAIN CANADA ISRAEL SUDAN CEYLON ITALY SWEDEN CHILE IVORY COAST SWITZERLAND CHINA JAMAICA SYRIAN ARAB REPUBLIC COLOMBIA JAPAN THAILAND CONGO, DEMOCRATIC JORDAN TUNISIA REPUBLIC OF KENYA TURKEY COSTA RICA KOREA, REPUBLIC OF UKRAINIA N SOVIET SOCIALIST CUBA KUWAIT REPUBLIC CYPRUS LEBANON UNION OF SOVIET SOCIALIST CZECHOSLOVAK SOCIALIST LIBERIA REPUBUCS REPUBLIC LIBYA UNITED ARAB REPUBLIC DENMARK LUXEMBOURG UNITED KINGDOM OF GREAT DOMINICAN REPUBLIC MADAGASCAR BRITAIN AND NORTHERN ECUADOR MALI IRELAND EL SALVADOR MEXICO UNITED STATES OF AMERICA ETHIOPIA MONACO URUGUAY FINLAND MOROCCO VENEZUELA FRANCE NETHERLANDS VIET-NAM NEW ZEALAND YUGOSLAVIA The Agency's Statute was approved on 26 October 1956 by the Conference -on the Statute of the IAEA held at United Nations Headquarters, New York; it entered into force on 29 July 1951. The Headquarters of the Agency are situated in Vienna. Its principal objective is "to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world". © IAEA, 1967 Permission to reproduce or translate the information contained in this publication may be obtained by writing to the International Atomic Energy Agency, Kârntner Ring 11, A-1010 Vienna I, Austria.
    [Show full text]
  • Silicon Nitride, a Close to Ideal Ceramic Material for Medical Application
    ceramics Review Silicon Nitride, a Close to Ideal Ceramic Material for Medical Application Robert B. Heimann Am Stadtpark 2A, D-02826 Görlitz, Germany; [email protected]; Tel.: +49-3581-667851 Abstract: This topical review describes the salient results of recent research on silicon nitride, a ceramic material with unique properties. The outcome of this ongoing research strongly encourages the use of monolithic silicon nitride and coatings as contemporary and future biomaterial for a variety of medical applications. Crystallographic structure, the synthesis and processing of monolithic structures and coatings, as well as examples of their medical applications that relate to spinal, orthopedic and dental implants, bone grafts and scaffolds, platforms for intelligent synthetic neural circuits, antibacterial and antiviral particles and coatings, optical biosensors, and nano-photonic waveguides for sophisticated medical diagnostic devices are all covered in the research reviewed herein. The examples provided convincingly show that silicon nitride is destined to become a leader to replace titanium and other entrenched biomaterials in many fields of medicine. Keywords: silicon nitride; structure; properties; processing; coatings; spinal implants; arthroplastic implants; bone scaffolds; dental implants; neural circuits; biosensors; medical diagnostics Citation: Heimann, R.B. Silicon 1. Introduction Nitride, a Close to Ideal Ceramic Material for Medical Application. Today, the research and development of metallic, ceramic, polymeric and composite Ceramics 2021, 4, 208–223. biomaterials have progressed to a high level of involvement and sophistication. This is https://doi.org/10.3390/ related to the fact that an ever-increasing part of the world population is in need of the repair ceramics4020016 or replacement of dysfunctional or damaged parts of the body, such as dental roots and teeth, alveolar ridge augmentation, intraocular lenses, heart pacemakers, cochlear implants, Academic Editors: Stuart Hampshire and hip and knee endoprostheses [1].
    [Show full text]