Determination of Zirconium Isotope Composition and Concentration For
Total Page:16
File Type:pdf, Size:1020Kb
Load more
Recommended publications
-
Coupled Thermomechanical Responses of Zirconium Alloy System Claddings Under Neutron Irradiation
applied sciences Article Coupled Thermomechanical Responses of Zirconium Alloy System Claddings under Neutron Irradiation Hui Zhao 1,†, Chong Yang 1,†, Dongxu Guo 1, Lu Wu 2, Jianjun Mao 2, Rongjian Pan 2, Jiantao Qin 2 and Baodong Shi 1,* 1 National Engineering Research Center for Equipment and Technology of Cold Rolled Strip, School of Mechanical Engineering, Yanshan University, Qinhuangdao 066004, China; [email protected] (H.Z.); [email protected] (C.Y.); [email protected] (D.G.) 2 State Key Laboratory of Nuclear Fuel and Materials, The First Sub-Institute, Nuclear Power Institute of China, Chengdu 610041, China; [email protected] (L.W.); [email protected] (J.M.); [email protected] (R.P.); [email protected] (J.Q.) * Correspondence: [email protected]; Tel.: +86-33-5838-7652 † H.Z. and C.Y. contributed equally to this work. Abstract: Zirconium (Zr) alloy is a promising fuel cladding material used widely in nuclear reactors. Usually, it is in service for a long time under the effects of neutron radiation with high temperature and high pressure, which results in thermomechanical coupling behavior during the service process. Focusing on the UO2/Zr fuel elements, the macroscopic thermomechanical coupling responses of pure Zr, Zr-Sn, and Zr-Nb binary system alloys, as well as Zr-Sn-Nb ternary system alloy as cladding materials, were studied under neutron irradiation. As a heat source, the thermal conductivity and thermal expansion coefficient models of the UO2 core were established, and an irradiation growth model of a pure Zr and Zr alloy multisystem was built. -
Production of Radionuclides
CHAPTER 15 Production of Radionuclides Contents 15.1. General considerations 389 15.2. Irradiation yields 390 15.3. Second-order reactions 393 15.4. Target considerations 397 15.4.1. Physical properties 397 15.4.2. Chemical properties 398 15.5. Product specifications 399 15.5.1. Radiochemical processing 399 15.5.2. Specific activity 400 15.5.3. Labeling 400 15.5.4. Radiochemical purity 401 15.6. Recoil separations 402 15.6.1. Target recoil products 402 15.6.2. Hot atom reactions 403 15.6.3. The Szilard-Chalmers process 404 15.7. Fast radiochemical separations 406 15.7.1. Production of 11C labeled compounds 407 15.7.2. Auto-batch procedures 407 15.7.3. On-line procedures: gas-phase separation 409 15.7.4. On-line procedures: solvent extraction 412 15.7.5. Mass separator procedures 412 15.8. Exercises 412 15.9. Literature 414 This chapter discusses production of radionuclides for beneficial use in science, medicine and technology. The nuclear fundamentals for the production processes have been given in Chapters 11 to 14. The formation of radionuclides is discussed in several Chapters: e.g. cosmogenic reactions leading to the formation of short-lived radionuclides in nature (Ch. 5 and 10); thermonuclear reactions leading to the formation of long-lived radioactivity in the universe (Ch. 17); the synthesis of trans-uranium elements (Ch. 16 and 19-21). The production and isolation of separated fission products is treated separately (Ch. 19-21). This chapter discusses aspects of fundamental importance to the production of radionuclides by a variety of methods. -
Electromagnetic Evaluation of Hydrogen Content in Zr 2.5%Nb Alloy
Electromagnetic evaluation of hydrogen content in Zr 2.5%Nb alloy R.Grimberg 1, M.L.Craus 1,2 , A.Savin 1, R.Steigmann 2, M.Ruch 3 1 National Institute of Research and Development for Technical Physics, Iasi, Romania 2 Joint Institute for Nuclear Research, Dubna, Russia 3 CNEA, Buenos Aires, Argentina Abstract Zr 2.5%Nb alloys have wide utilization in nuclear energy engineering, being used at the fabrication of pressure tubes from Pressurized Heavy Water Reactors (PHWR). This paper proposes to present an electromagnetic method for determination of diffused hydrogen/deuterium content based on the measurement of conductivity, correlating the theoretical results with those experimental obtained on samples cropped from non-irradiated pressure tubes. The methods can be used directly, the results being obtained by data inversion. Keywords: PHWR, hydrogen content, electromagnetic nondestructive evaluation method, conductivity 1. Introduction Zirconium alloys represents the majority of structural materials from nuclear reactors, Light Water Reactors types (LWR) and Pressurized Heavy Water reactor” (PHWR). Zirconium alloys are exclusively used at the manufacturing of nuclear fuel cladding, nuclear fuel bundle subassemblies and fuel channels. During the reactor functioning, the zirconium alloy absorbs hydrogen or deuterium that can appear due to the radiolysis of water used as coolant agent or due to heavy water according to the reactions → + HO2 H 2 O (a) → + D2 O D 2 O (b) Also, hydrogen and deuterium can appear due to oxidation of zirconium alloys + → + Zr 2H2 O ZrO 2 2H 2 (c) + → + Zr 2D2 O ZrO 2 2D 2 (d) Hydrogen in solid solution diffuses due to the existence of concentration, temperature and stress gradients. -
Evaluate the Contribution of the Fuel Cladding Oxidation Process on The
Evaluate the contribution of the fuel cladding oxidation process on the hydrogen production from the reflooding during a potential severe accident in a nuclear reactor Florian Haurais To cite this version: Florian Haurais. Evaluate the contribution of the fuel cladding oxidation process on the hydrogen production from the reflooding during a potential severe accident in a nuclear reactor. Materials. Université Paris-Saclay, 2016. English. NNT : 2016SACLS375. tel-01519349 HAL Id: tel-01519349 https://tel.archives-ouvertes.fr/tel-01519349 Submitted on 6 May 2017 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. NNT : 2016SACLS375 THÈSE DE DOCTORAT DE L’UNIVERSITÉ PARIS-SACLAY PRÉPARÉE À L’UNIVERSITÉ PARIS-SUD ÉCOLE DOCTORALE N°576 Particules Hadrons Énergie et Noyau : Instrumentation, Image, Cosmos et Simulation Spécialité de doctorat : Énergie Nucléaire Par M. Florian Haurais Evaluate the contribution of the fuel cladding oxidation process on the hydrogen production from the reflooding during a potential severe accident in a nuclear reactor Thèse présentée et soutenue à Palaiseau, le Lundi 14 Novembre 2016. Composition du Jury : M. Frédérico Garrido, Professeur des universités, Université Paris-Sud, Président du jury M. Arthur Motta, Professeur, Pennsylvania State University, Rapporteur M. -
5.03 Corrosion of Zirconium Alloys
This article was originally published in the Comprehensive Nuclear Materials published by Elsevier, and the attached copy is provided by Elsevier for the author's benefit and for the benefit of the author's institution, for non-commercial research and educational use including without limitation use in instruction at your institution, sending it to specific colleagues who you know, and providing a copy to your institution’s administrator. All other uses, reproduction and distribution, including without limitation commercial reprints, selling or licensing copies or access, or posting on open internet sites, your personal or institution’s website or repository, are prohibited. For exceptions, permission may be sought for such use through Elsevier's permissions site at: http://www.elsevier.com/locate/permissionusematerial Allen T.R., Konings R.J.M., and Motta A.T. (2012) Corrosion of Zirconium Alloys. In: Konings R.J.M., (ed.) Comprehensive Nuclear Materials, volume 5, pp. 49-68 Amsterdam: Elsevier. © 2012 Elsevier Ltd. All rights reserved. Author's personal copy 5.03 Corrosion of Zirconium Alloys T. R. Allen University of Wisconsin, Madison, WI, USA R. J. M. Konings European Commission, Joint Research Centre, Institute for Transuranium Elements, Karlsruhe, Germany A. T. Motta The Pennsylvania State University, University Park, PA, USA ß 2012 Elsevier Ltd. All rights reserved. 5.03.1 Introduction 49 5.03.2 General Considerations 50 5.03.2.1 Oxidation 50 5.03.2.2 Hydrogen Uptake 51 5.03.2.3 Controlling Factors for Corrosion 52 5.03.3 Uniform -
High Temperature Oxidation of Zirconium Base Alloy in Steam
HIGH TEMPERATURE OXDATION OF A ZIRCONIUM BASE ALLOY IN STEAM KWANGHEON PARK"), TAEGEUN YOO', SUNGKWONKIM') HYUN-GlL KIM2 YONGHWAN JEONG2I, KYUTAE KIM3 ) 1)=ghee University, South Korea 2\Comra Atomic Energy Research Institute 3)KEPCO Nuclear Fuel Company, South Korea Abstract High temperature steam oxidation behaviors of a Zirconium alloy, Zr-1%oNb alloy was examined for the comparison to those of Zircaloy-4 (Zry-4). Testing temperatures were 700 - 12000C. At atmospheric steam pressure, oxidation kinetics of Zr-lNb alloy follows parabolic-rate law, instead of cubic-rate as was observed in Zry-4 below 900 C. The oxidation rate ofZr-lNb alloy is slightly lower than that ofZry-4. A double layer autoclave, that can make high steam pressures up to 50bar and temperatures up to 900 0C, was used to get the steam pressure effects on high temperature oxidation. Zry-4 was very sensitive to the steam pressure, and the oxidation rate increases exponentially with applied steam pressure. Zr-IlNb alloy was less sensitive to the high-pressure steam. The enhancement parameter is about 3 to 13 times lower than that of Zry-4. The stability oftetragonal phase in the Zr-I %oNb alloy comparing Zry-4 seems to make the differcnue in oxidation kinetics. 1. INTRODUCTION Zr-base alloys are used as cladding materials for nuclear fuel in light and heavy water reactors. Zricaloy-4 (Zry-4) has been used satisfactorily as a cladding material in pressurized water reactors. Nowadays, light water reactors tend to extend their fuel cycle length with high bum-up of nuclear fuel to get the improved economy. -
Surface Chemistry of Zirconium
Progress in Surface Science 78 (2005) 101–184 www.elsevier.com/locate/progsurf Review Surface chemistry of zirconium N. Stojilovic, E.T. Bender, R.D. Ramsier * Departments of Physics and Chemistry, The University of Akron, 250 Buchtel Commons, Ayer Hall 111, Akron, OH 44325-4001, USA Abstract This article presents an overview of the surface chemistry of zirconium, focusingon the relationship of what is known from model studies and how this connects to current and future applications of Zr-based materials. The discussion includes the synergistic nature of adsorbate interactions in this system, the role of impurities and alloyingelements, and temperature- dependent surface–subsurface transport. Finally, some potential uses of zirconium and its alloys for biomedical and nanolithographic applications are presented. Ó 2005 Elsevier Ltd. All rights reserved. Keywords: Zirconium; Oxidation; Surface chemistry; Subsurface species; Diffusion; Water; Oxygen; Hydrogen; Nuclear materials; Alloys; Zircaloy Contents 1. Contextual overview........................................ 102 2. Systems of interest ......................................... 104 2.1. Water ............................................. 104 2.2. Oxygen ............................................ 122 2.3. Hydrogen .......................................... 132 2.4. Sulfur ............................................. 143 2.5. Carbon ............................................ 147 * Correspondingauthor. Tel.: +1 330 9724936; fax: +1 330 9726918. E-mail address: [email protected] (R.D. -
Lawrence Berkeley National Laboratory Recent Work
Lawrence Berkeley National Laboratory Recent Work Title RADIOCHEMICAL AND SPECTROMETER STUDIES OF SEVERAL NEUTRON-DEFICIENT ZIRCONIUM ISOTOPES AND THEIR DECAY PRODUCTS Permalink https://escholarship.org/uc/item/5ck8q7xw Authors Hyde, Earl K. O'Kelley, Grover D. Publication Date 1950-12-28 eScholarship.org Powered by the California Digital Library University of California UCRL- .Jt)AJ,-.64.. _.; >-w .J w ... ~ ,• 0:: w m I . ;•· <( z 0:: 0 LL ..J u<( LL 0 >- 1-- TWO-WEEK LOAN COPY (/) This is a Library Circulating Copy 0:: which may be borrowed for two weeks. w For a personal retention copy, call > Tech. Info. Diuision, Ext. 5545 z '.-o:-. :J ~~ RADIATION LABORATORY DISCLAIMER This document was prepared as an account of work sponsored by the United States Government. While this document is believed to contain correct information, neither the United States Government nor any agency thereof, nor the Regents of the University of California, nor any of their employees, makes any warranty, express or implied, or assumes any legal responsibility for the accuracy, completeness, or usefulness of any · information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by its trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof, or the Regents of the University of California. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof or the Regents of the University of California. -
Technische Universität München Physik-Department Study of The
Technische Universität München Physik-Department Study of the biokinetics of zirconium isotopes in humans and its relevance to internal dosimetry Matthias Greiter Vollständiger Abdruck der von der Fakultät für Physik der Technischen Universität München zur Erlangung des akademischen Grades eines Doktors der Naturwissenschaften genehmigten Dissertation. Vorsitzender: Univ.-Prof. Dr. J. L. van Hemmen Prüfer der Dissertation: 1. Hon.-Prof. Dr. H.G. Paretzke 2. Univ.-Prof. Dr. F. von Feilitzsch Die Dissertation wurde am 27.09.2007 bei der Technischen Universität München eingereicht und durch die Fakultät für Physik am 22.04.2008 angenommen. i Author contact information: Matthias Greiter Institute of Radiation Protection Helmholtz Zentrum München German Research Center for Environmental Health (GmbH) Ingolstaedter Landstrasse 1 85764 Neuherberg Germany E-mail: [email protected] ii Table of contents Abstract ...................................................................................................................................... 1 List of acronyms, symbols and abbreviations ............................................................................ 2 1 Aim of the study......................................................................................................... 3 1.1 Natural occurrence of zirconium compounds ............................................................ 3 1.2 Applications of stable zirconium................................................................................ 5 1.3 Applications -
Effect of Alloying Elements on Hydrogen Pickup in Zirconium Alloys
ZIRCONIUM IN THE NUCLEAR INDUSTRY: 17TH INTERNATIONAL SYMPOSIUM 479 STP 1543, 2014 / available online at www.astm.org / doi: 10.1520/STP154320120215 Adrien Couet,1 Arthur T. Motta,2 and Robert J. Comstock3 Effect of Alloying Elements on Hydrogen Pickup in Zirconium Alloys Reference Couet, Adrien, Motta, Arthur T., and Comstock, Robert J., “Effect of Alloying Elements on Hydrogen Pickup in Zirconium Alloys,” Zirconium in the Nuclear Industry: 17th International Symposium, STP 1543, Robert Comstock and Pierre Barberis, Eds., pp. 479–509, doi:10.1520/ STP154320120215, ASTM International, West Conshohocken, PA 2014.4 Abstract Although the optimization of zirconium-based alloys has led to significant improvements in hydrogen pickup and corrosion resistance, the mechanisms by which such alloy improvements occur are still not well understood. In an effort to understand such mechanisms, we conducted a systematic study of the alloy effect on hydrogen pickup, using advanced characterization techniques to rationalize precise measurements of hydrogen pickup. The hydrogen pickup fraction was accurately measured for a specially designed set of commercial and model alloys to investigate the effects of alloying elements, microstructure, and corrosion kinetics on hydrogen uptake. Two different techniques for measuring hydrogen concentrations were used: a destructive technique, vacuum hot extraction, and a non-destructive one, cold neutron prompt gamma activation analysis. The results indicate that hydrogen pickup varies not only from alloy to alloy, but also during the corrosion process for a given alloy. These variations result from the process of charge balance during the corrosion reaction, such that the pickup of hydrogen decreases when the rate of electron transport or Manuscript received December 25, 2012; accepted for publication June 26, 2013; published online June 17, 2014. -
Towards an Understanding of Zirconium Alloy Corrosion
v V.. AECL-5548 ATOMIC ENERGY L'ENERGIE ATOMIQUE OF CANADA LIMITED DU CANADA LIMITEE TOWARDS AN UNDERSTANDING OF ZIRCONIUM ALLOY CORROSION by B. COX Presented on the occasion of the award of the William J, Kroll Medal, Quebec City, P.Q. August 1976 Chalk River Nuclear Laboratories Chalk River, Ontario August 1976 Cover Photograph: Large columnar oxide grains on zirconium oxidised in air at 650°C, magnification x20,000. TOWARDS AW UNDERSTANDING OF ZIRCONIUM ALLOY CORROSION by B. Cox (M.A., Ph.D., Cantab.) Head of Materials Science Branch Atomic Energy of Canada Limited Chalk River Nuclear Laboratories Chalk River, Ontario KOJ 1J0 on the occaiZon o & the William J. Kroll Medal Presentation Quebec City, P.Q. August 11, 19 76 AECL-5548 ^^mjrrejid_re_lji corrosion des alliages de zi rconium par B. Cox à l'occasion de la présentation de la Médaille William J. Krol 1 Québec, P.Q. 11 août 1976 Résumé On donne un bref historique du développement d'un programme visant à mieux comprendre les mécanismes de corrosion qui jouent dans les alliages de zirconium. Un sommaire général indique les progrès réalisés jusqu'à présent dans la mise en oeuvre de ce programme. L'Energie Atomique du Canada, Limitée Laboratoires Nucléaires de Chalk River Chalk River, Ontario AoQt 1976 AECL-5548 TOWARDS AN UNDERSTANDING OF ZIRCONIUM ALLOY CORROSION by B . Cox on the occasion of th(. William J. Kroll Medal Presentation Quebec City, P.Q. August 11, 1976 ABSTRACT A brief historical summary is given of the development of a programme for understanding the corrosion mechanisms operating for zirconium alloys. -
Zirconium Alloys in Nuclear Technology
Proc. Indian Aead. Sci. (Engg. Sei.) Vol. 4, Pt. 1, April 1981: pp. 41-56. ~) Printed in India. Zirconium alloys in nuclear technology R KRISHNAN and M K ASUNDI Metallurgy Division, Bhabha Atomic Research Centre, Trombay, Bombay 400 085 MS received 17 October 1980 Abstract. This paper describes the historical development of zirconium and its alloys as structural materials for nuclear reactors. The various problems encountered in the early stages of the development of zircaloys and their performance in reactors operating presently are described in detail. The development of Zr-2.5~o Nb alloys for pressure tube applications is discussed. The paper concludes with a detailed dis- cussion on the development potential of zirconium alloys for high temperature appli- cations and a brief account of the work carried out at Trombay in this field. Keywords. Zirconium; zirconium alloys; structural materials; nuclear reactors; corrosion; high strength alloys. 1. Introduction The development of zirconium metallurgy is essentially due to the nuclear industry, where zirconium alloys have come to be regarded as the proven structural material. This is primarily because of their unique combination of good corrosion resistance in water near 300~ and low capture cross-section for thermal neutrons (Douglass 1971). It is quite likely that the application of zirconium in the nuclear industry will remain its dominant use. This paper begins with an account of the present day use of zirconium alloys in the nuclear industry mainly to acquaint non-nuclear techno- logists with the various sizes, shapes and functions of such structural materials. Then the history of the development of presently accepted zirconium alloys--zirca- loy-2 and zircaloy-4---is given some consideration.