DOCSLIB.ORG

Irradiation Performance of Fast Reactor Fuels

Total Page：16

File Type：pdf, Size：1020Kb

Download full-text PDF Read full-text
Irradiation Performance of Fast Reactor Fuels Oktober 1967 KFK 662 EUR 3698 e Institut für Angewandte Reaktorphysik Institut für Material- und Festkörperforschung Irradiation Performance of Fast Reactor Fuels D. Geithoff, G. Karsten, K. Kummerer GESELLSCHAFT FUR KERNF( KARLSRUHE Als Manuskript vervielfältigt Für diesen Bericht behalten wir uns alle Rechte vor Gesellschaft für 'Kernforschung m.b.H. Karlsruhe KERNFORSCHUNGSZENTRUM KARLSRUHE Oktober I967 KFK­662 EUR 3698 e Institut für Angewandte Reaktorphysik Institut für Material­ und Festkörperforschung IRRADIATION PERFORMANCE OF FAST REACTOR FUELS *' D. Geithoff, G. Karsten and Κ. Kümmerer Paper presented at the Symposium on Plutonium Fuels Technology, Nuclear Metallurgy Committee IMD­AIME, Oct.4­6,1967, Phoenix, Ariz., USA. Gesellschaft für Kernforschung mbH., Karlsruhe *) Work performed within the association in the field of fast reactors between the European Atomic Energy Community and Gesellschaft für Kernforschung mbH., Karlsruhe ABSTRACT The requirements for uranium-plutonium dioxide fuels are the basis for a program of irradiation tests which is outlined in the introduction. Some hypothetical considerations supply a rational background for the design of specimens and the chosen operation conditions. The experimental facilities and irradiation devices are shortly described and discussed with respect to the irradiation characteristics. The available hot cell equipment provides for the standard examination technique including X-raying by betatron and fission gas chromatography. One group of experiments refers to the short term behaviour of the oxide type fuel. In another group the long time behaviour of oxide fuel pins is studied. Different pins have been irradiated in capsules and in a helium loop. The present results indicate, that the basic assumptions for the pin design are principally con­ firmed, the specimen failures only arising from malfunctions of the irradiation devices. The now available irradiation experience is used for the fabrication of further irradiation specimens of more realistic fast reactor dimensions and also for the concept of fast breeder pin specifications. Under these aspects the forth­ coming main experiments are discussed. Finally the status of knowl­ edge and experimental experience is composed in a set of conclusive remarks. INTRODUCTION The operational requirements for oxide type fuel, which are 1 described in full details in the related paper in the first session of this symposium, are the principal guidelines for all irradiation test work in the Karlsruhe fast breeder project. The most important technical features are the proper irradiation conditions with res­ pect to linear rod power, to cladding temperature and to the ex­ pected total burnup performance. The typical external and internal geometry is characterized by a small fuel diameter (compared to thermal reactor fuel) and a length distribution which includes an ι active fuel zone, at least one axial blanket region and a fission gas plenum. The typical ranges of requirements both for sodium and · steam cooled pins are compiled in Table I. The whole irradiation program is oriented towards these target data. As in the first period only a thermal reactor with a limited neutron flux, the Karlsruhe research reactor FR2, was available, significantly larger fuel diameters were necessary in order to reach the linear rod power wanted. Above the intention of a mere perform• ance test up to high burnups - a goal which will be achieved by tests currently underway - the here described work refers partly to parameter studies. The short term tests try to evaluate a system• atic scheme of parameter variations. The quoted long term tests are only a first step, which has to demonstrate the feasibility of the irradiation equipment and furthermore to experience the post irradia• tion examination routine. The next step with much higher target burnups and parameter constellations more similar to fast reactor conditions are already in operation, on the one side with improved irradiation devices in the FR2 again, on the other side with fast flux irradiations in the Dounreay Fast Reactor. The present paper describes the status of our experimental potential and experience and also some introductory theoretical considerations. We feel it necessary, to establish simple model assumptions concerning the expected irradiation behaviour as a rough guide. IRRADIATION HYPOTHESES The target burnup for a fast breeder fuel pin is 100 000 MWd/t. In order to make predictions about the pin behaviour with• in an equivalent irradiation time, the pin design is based on a theoretical fuel pin behaviour analysis. In this context two limi• tating assumptions are made, namely - the fuel operation time will come to an and, after the avail• able void volume of the most critical fuel cross section has been filled by fission-induced fuel swelling to the highest possible extent, ­ the fuel operation might end earlier by mechanical contacts between cladding and non­plastic fuel. As a consequence of these assumptions three separate questions can be discussed with theoretical methods: 1. What is the maximum achievable burnup locally, if only the volume swelling due to fission products is considered? 2. Applying the answer of question 1, where is the most critical fuel cross section along the fuel pin located? 3. Does a socalled "non­plastic" fuel behaviour arise under the planned operation conditions, leading to a mechanical attack on the cladding? In order to produce an answer to the first question a burnup model is established using some schematic considerations: ­ The internal volume of the pin consists of the fuel volume Vp, F' the (hot) gap volume between fuel and cladding V«, and the dishing volume (if applicable) V_. ­ The fuel volume itself can be divided into a plastic zone V ­ with temperatures above 17OO C, a creep zone V with tempe­ vi ratures between I3OO and I7OO C, where creep and diffusion are of remarkable velocity, and a low temperature zone V.. O TE il "•■t below 1300 C with visco­elastic behaviour mainly ' ' . ­ Furthermore there are defined availability factors £,, m and n in the 3 fuel zones, which give the volume portions of the porosity being actually available for volume expansion due to swelling processes. The porous volume should be available to any expansion with about 80 v/o above I7OO C, with 50 v/o at 13OO­I7OO C and ­ according to a numerical evaluation with not more than 30 v/o below 1300°C. The last figure may rise somewhat during long time operation as the fuel becomes more plastic at low temperatures. Hence we apply for numerical evaluations L = 0.8, m = O.5, n = 0.3· ­ Finally two swelling processes are envisaged which need a con­ tinuously increasing amount of space with burnup. The basic swelling effect is defined here to be caused by solid fission products and a contribution of fission gases, working in all 5 three zones. According to a literature evaluation this basic swelling rate h seems to be 1.6 % volume change in 100 % S dense oxide per 10 MWd/kg heavy metal burnup. That means a figure h =1.6x10 kg/MWd. An additional swelling effect S caused by fission gases mainly is added for the creep zone. Its rate is assumed to be h = 0.4 χ 10— 3 kg/MWd. With these explanations the maximum possible burnup Β (in MWd per kg heavy metal content), after which all available volume is used up by the swelling fuel, is: (gV + mV + IlV ) +V + V Β „ °-V pl cr lt G D (<|) ^Fhs VF + hg Vcr where ,$_ means the fuel density in the pellet fuel. A graphic evaluation of this burnup formula is given in Fig.1 for a linear rod power of 500 W/cm. The second question for the most critical cross section is handled by applying the burnup formula, which represents the fuel swelling capacity. Hence locally the maximum allowable burnup can be evaluated for the axial power distribution of a pin. These local maximum burnup figures have to be compared with the really arising burnup at each axial part of a pin in equal times. For certain parts of a pin, which are not necessarily the absolute maximum burnup ranges, the capacity for burnup is implemented firstly. These constitute the critical fuel cross sections. Such an analysis can be expanded of course on a total reactor core by application of specific thermal design data. Generally the critical zones are in the half of the core with the coolant inlet. To deal with the third question for the existence of non- plastic fuel, the expansion rate of the fuel, which is given by the swelling rate, is compared to a compression rate, which is produced by an evaluation of fuel-cladding interaction, Fig.2. Here irradiation damages and thermal stresses with the cladding are taken into account by a pessimistic assumption for the 0.2 $-creep stress for 10 000 hours of the considered stainless steel. 2 3 If Furthermore, by application of mechanical oxide data ' ' the force of the cladding on the swelling fuel ring with the volume V It is calculated. The result is, that, under the assumption of 950 C average ring temperature, only those parts of the fuel with linear rod powers lower than 250 W/cm behave non-plastic. They are situat• ed beyond the critical zones mentioned above and hence do not re• present limitations for operation time in such designs. EXPERIMENTAL FACILITIES The irradiation program in the Karlsruhe research reactor FR2 a heavy water-moderated reactor with kk MW thermal power - is performed with two types of irradiation equipment. There is a helium loop operating in the central channel and furthermore, there can be introduced irradiation capsules into normal fuel sub• assembly positions. The helium loop can be loaded with two diffe• rent irradiation inserts. For long term irradiation an insert is used containing k fuel pin specimens in a cluster.
Load more

Recommended publications
  • Institute for Clinical and Economic Review
    INSTITUTE FOR CLINICAL AND ECONOMIC REVIEW FINAL APPRAISAL DOCUMENT BRACHYTHERAPY & PROTON BEAM THERAPY FOR TREATMENT OF CLINICALLY-LOCALIZED, LOW-RISK PROSTATE CANCER December 22, 2008 Senior Staff Daniel A. Ollendorf, MPH, ARM Chief Review Officer Julia Hayes, MD Lead Decision Scientist Pamela McMahon, PhD Sr. Decision Scientist Steven D. Pearson, MD, MSc President, ICER Associate Staff Michelle Kuba, MPH Sr. Technology Analyst Angela Tramontano, MPH Research Assistant © ICER, 2008 1 CONTENTS About ICER .................................................................................................................................. 3 Acknowledgments ...................................................................................................................... 4 Executive Summary .................................................................................................................... 5 Evidence Review Group Deliberation.................................................................................. 15 ICER Integrated Evidence Rating.......................................................................................... 21 Evidence Review Group Members........................................................................................ 24 Appraisal Overview.................................................................................................................. 28 Background ...............................................................................................................................
    [Show full text]
  • 小型飛翔体／海外 ［Format 2］ Technical Catalog Category
    小型飛翔体／海外 ［Format 2］ Technical Catalog Category Airborne contamination sensor Title Depth Evaluation of Entrained Products (DEEP) Proposed by Create Technologies Ltd & Costain Group PLC １．DEEP is a sensor analysis software for analysing contamination. DEEP can distinguish between surface contamination and internal / absorbed contamination. The software measures contamination depth by analysing distortions in the gamma spectrum. The method can be applied to data gathered using any spectrometer. Because DEEP provides a means of discriminating surface contamination from other radiation sources, DEEP can be used to provide an estimate of surface contamination without physical sampling. DEEP is a real-time method which enables the user to generate a large number of rapid contamination assessments- this data is complementary to physical samples, providing a sound basis for extrapolation from point samples. It also helps identify anomalies enabling targeted sampling startegies. DEEP is compatible with small airborne spectrometer/ processor combinations, such as that proposed by the ARM-U project – please refer to the ARM-U proposal for more details of the air vehicle. Figure 1: DEEP system core components are small, light, low power and can be integrated via USB, serial or Ethernet interfaces. 小型飛翔体／海外 Figure 2: DEEP prototype software ２．Past experience (plants in Japan, overseas plant, applications in other industries, etc) Create technologies is a specialist R&D firm with a focus on imaging and sensing in the nuclear industry. Createc has developed and delivered several novel nuclear technologies, including the N-Visage gamma camera system. Costainis a leading UK construction and civil engineering firm with almost 150 years of history.
    [Show full text]
  • A 50-100 Kwe Gas-Cooled Reactor for Use on Mars
    SANDIA REPORT SAND2006-2189 Unlimited Release Printed April 2006 A 50-100 kWe Gas-cooled Reactor For Use On Mars Curtis D. Peters Prepared by Sandia National Laboratories Albuquerque, New Mexico 87185 and Livermore, California 94550 Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy’s National Nuclear Security Administration under Contract DE-AC04-94AL85000. Approved for public release; further dissemination unlimited. Issued by Sandia National Laboratories, operated for the United States Department of Energy by Sandia Corporation. NOTICE: This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government, nor any agency thereof, nor any of their employees, nor any of their contractors, subcontractors, or their employees, make any warranty, express or implied, or assume any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represent that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government, any agency thereof, or any of their contractors or subcontractors. The views and opinions expressed herein do not necessarily state or reflect those of the United States Government, any agency thereof, or any of their contractors. Printed in the United States of America. This report has been reproduced directly from the best available copy. Available to DOE and DOE contractors from U.S.
    [Show full text]
  • Advances in Small Modular Reactor Technology Developments
    Advances in Small Modular Reactor Technology Developments Advances in Small Modular Reactor Technology Developments Technology in Small Modular Reactor Advances A Supplement to: IAEA Advanced Reactors Information System (ARIS) 2018 Edition For further information: Nuclear Power Technology Development Section (NPTDS) Division of Nuclear Power IAEA Department of Nuclear Energy International Atomic Energy Agency Vienna International Centre PO Box 100 1400 Vienna, Austria Telephone: +43 1 2600-0 Fax: +43 1 2600-7 Email: [email protected] Internet: http://www.iaea.org Printed by IAEA in Austria September 2018 18-02989E ADVANCES IN SMALL MODULAR REACTOR TECHNOLOGY DEVELOPMENTS 2018 Edition A Supplement to: IAEA Advanced Reactors Information System (ARIS) http://aris.iaea.org DISCLAIMER This is not an official IAEA publication. The material has not undergone an official review by the IAEA. The views expressed do not necessarily reflect those of the International Atomic Energy Agency or its Member States and remain the responsibility of the contributors. Although great care has been taken to maintain the accuracy of information contained in this publication, neither the IAEA nor its Member States assume any responsibility for consequences which may arise from its use. The use of particular designations of countries or territories does not imply any judgement by the publisher, the IAEA, as to the legal status of such countries or territories, of their authorities and institutions or of the delimitation of their boundaries. The mention of names of specific companies or products (whether or not indicated as registered) does not imply any intention to infringe proprietary rights, nor should it be construed as an endorsement or recommendation on the part of the IAEA.
    [Show full text]
  • Self-Sustaining Thorium Boiling Water Reactors
    Project No. 11-3023 Self-Sustaining Thorium Boiling Water Reactors Reactor Concepts Dr. Ehud Greenspan University of California-Berkeley In collaboraon with: Massachuses Instute of Technology University of Michigan Brookhaven Naonal Laboratory Brian Robinson, Federal POC Temitope Taiwo, Technical POC NEUP Project # 11-3023 September 2011 to December 2014 Self-sustaining thorium boiling water reactors Summary Report Ehud Greenspan, Phillip M. Gorman, Sandra Bogetic, Jeffrey E. Seifried, Guanheng Zhang, Christopher R. Varela, Massimiliano Fratoni and Jasmina L. Vujic UC Berkeley, [email protected] Thomas Downar, Andrew Hall, Andrew Ward, Michael Jarrett, Aaron Wysocki and Yunlin Xu University of Michigan, [email protected] Mujid Kazimi, Koroush Shirvan and Alexander Mieloszyk MIT, [email protected] Michael Todosow, Nicholas Brown and Lap Cheng BNL, [email protected] 28 February 2015 1 Table of Contents Overview Page Executive summary 3 1. Introduction 7 2. The RBWR cores 7 2.1 Hitachi RBWR cores 7 2.2 Independent evaluation of the Hitachi designs 14 2.3 The incentive for thorium-based RBWR 15 2.4 The thorium-based RBWR core designs 17 3. Development of computational methods for RBWR type cores 19 3.1 3-D core simulator 20 3.2 Monte-Carlo based coupled neutronics – TH – depletion analysis capability 21 3.3 Stability and safety analysis capability 21 3.4 Fuel performance analysis capability 22 3.5 Thermal hydraulic correlations for tight lattice high void cores 24 4. Feasibility of fuel-self-sustaining RBWR-Th cores 24 4.1 Study strategy 24 4.2 RBWR-Th core designs 26 4.3 RBWR-Th safety and stability 30 4.4 RBWR-Th fuel performance 31 5.
    [Show full text]
  • Latest Fusion R&D Activities at INEST
    1 Latest Fusion R&D Activities at INEST Presented by Yican Wu (Zhibin Chen on behalf) Contributed by FDS Team Institute of Nuclear Energy Safety Technology (INEST) Chinese Academy of Sciences www.fds.org.cn FDS: Frontier Development of Science 2 Institute of Nuclear Energy Safety Technology Chinese Academy of Sciences ∙ FDS Team Main Research Areas Fundamental research: • Neutron Physics Applied research: 1. Advanced Fission Energy LFR etc. 2. Fusion Energy FNT & Materials etc. Headquarter in Hefei, Anhui, China 3. Extended Nuclear Technology Applications Key Programs: . Strategic Priority Research Program of CAS . ITER Related International and Domestic Program . Nuclear Energy & Safety Technology Innovation Program 400 Staff + 100 Students Three Integrated Test Reactors A professional institute is devoted to design and R&D of advanced nuclear energy and safety technology 3 Outline I. Highlights of Fusion R&D Activities • Fusion Neutron Sources • Neutronics Methodology and Simulation • Fusion Safety • Materials and TBM Technologies II. International Communication and Collaboration 4 1. Fusion Neutron Sources 5 Development of Fusion Neutron Sources at INEST ≥ 1018 n/ s 15 16 10 -10 n/s HINEG-IIIa 1012-1013 n/s Fusion Volumetric Neutron HINEG-IIa Source ( e.g. GDT-FVNS ) Mixed-beam HINEG-I Fusion Neutron Source (in Operation) . Fusion materials (≥20dpa/FPY) . Materials Irradiation . Multi-Physics Coupling Test . Nuclear Design and Performance . Nuclear Data . System Engineering Integration Test of Component Test . Methods and Software
    [Show full text]
  • Control Rod Safety Management in Nuclear Power Plant
    Journal of Advanced Review on Scientific Research 32, Issue 1 (2017) 14-27 Journal of Advanced Review on Scientific Penerbit Akademia Baru Research Journal homepage: www.akademiabaru.com/arsr.html ISSN: 2289-7887 Control rod safety m anage ment in nuclear power plant: A Open review Access Muhammad Adil Khattak 1, ∗, Filzah Hazirah Jaffar 1,Najaa Fadhilah Mohd Nasir 1, Nurfarhah 1 1 1 Ridzuan , Nurlaila Syamsul Bahri ,Wan Nur Iffah Haziqah Md Zamri 1 Department of Nuclear Engineering, Faculty of Chemical & Energy Engineering Universiti Teknologi Malaysia, 81310 Skudai Johor , Malaysia ARTICLE INFO ABSTRACT Article history: The primary objective of control rod management is to ensure the safe, reliable and Received 12 April 2017 optimum use of the nuclear fuel in the reactor, while remain within the limits imposed Received in revised form 24 May 2017 by the design of the fuel assembly and reactor w.r.t the safety analysis. In numerous Accepted 25 May 2017 reactors, the control rods perform the function of reactivity control, both globally and Available online 26 May 2017 locally, latter also control the power distribution of the core. Most control rods are completely withdrawn from the core during operations and fully inserted during shutdown of the reactor. This are the prove of an investigative study into optimization of the heterogeneous control rods, the safety management of an additional safety rod, and control rod drop hydrodynamic analysis which will be able to provide an efficient and maximum safety procedure for emergency shut down system in the reactor. It is also will be very important safety features of the reactor.
    [Show full text]
  • NRC Collection of Abbreviations
    I Nuclear Regulatory Commission c ElLc LI El LIL El, EEELIILE El ClV. El El, El1 ....... I -4 PI AVAILABILITY NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources: 1. The NRC Public Document Room, 2120 L Street, NW., Lower Level, Washington, DC 20555-0001 2. The Superintendent of Documents, U.S. Government Printing Office, P. 0. Box 37082, Washington, DC 20402-9328 3. The National Technical Information Service, Springfield, VA 22161-0002 Although the listing that follows represents the majority of documents cited in NRC publica- tions, it is not intended to be exhaustive. Referenced documents available for inspection and copying for a fee from the NRC Public Document Room include NRC correspondence and internal NRC memoranda; NRC bulletins, circulars, information notices, inspection and investigation notices; licensee event reports; vendor reports and correspondence; Commission papers; and applicant and licensee docu- ments and correspondence. The following documents in the NUREG series are available for purchase from the Government Printing Office: formal NRC staff and contractor reports, NRC-sponsored conference pro- ceedings, international agreement reports, grantee reports, and NRC booklets and bro- chures. Also available are regulatory guides, NRC regulations in the Code of Federal Regula- tions, and Nuclear Regulatory Commission Issuances. Documents available from the National Technical Information Service Include NUREG-series reports and technical reports prepared by other Federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission. Documents available from public and special technical libraries include all open literature items, such as books, journal articles, and transactions.
    [Show full text]
  • Understanding Non-Hodgkin Lymphoma
    Understanding Non-Hodgkin Lymphoma A Guide for Patients, Survivors, and Loved Ones July 2018 Lymphoma Research Foundation (LRF) Helpline and Clinical Trials Information Service CONTACT THE LRF HELPLINE Trained staff are available to answer questions and provide support to patients, caregivers and healthcare professionals in any language. Our support services include: • Information on lymphoma, treatment options, side effect management and current research fi ndings • Financial assistance for eligible patients and referrals for additional fi nancial, legal and insurance help • Clinical trial searches based on patient’s diagnosis and treatment history • Support through LRF’s Lymphoma Support Network, a national one-to one volunteer patient peer program Monday through Friday, Toll-Free (800) 500-9976 or email [email protected] Understanding Non-Hodgkin Lymphoma A Guide For Patients, Survivors, and Loved Ones July 2018 This guide is an educational resource compiled by the Lymphoma Research Foundation to provide general information on adult non- Hodgkin lymphoma. Publication of this information is not intended to replace individualized medical care or the advice of a patient’s doctor. Patients are strongly encouraged to talk to their doctors for complete information on how their disease should be diagnosed, treated, and followed. Before starting treatment, patients should discuss the potential benefits and side effects of cancer therapies with their physician. Contact the Lymphoma Research Foundation Helpline: (800) 500-9976 [email protected] Website: lymphoma.org This patient guide is supported through unrestricted educational grants from: © 2018 Lymphoma Research Foundation. Information contained herein is the property of the Lymphoma Research Foundation (LRF). Any portion may be reprinted or reproduced provided that LRF is acknowledged to be the source and the Foundation’s website (lymphoma.org) is included in the citation.
    [Show full text]
  • Fundamentals of Boiling Water Reactor (Bwr)
    FUNDAMENTALS OF BOILING In this light a general description of BWR is presented, with emphasis on the reactor physics aspects (first lecture), and a survey of methods applied in fuel WATER REACTOR (BWR) and core design and operation is presented in order to indicate the main features of the calculation^ tools (second lecture). The third and fourth lectures are devoted to a review of BWR design bases, S. BOZZOLA reactivity requirements, reactivity and power control, fuel loading patternc. Moreover, operating 'innts are reviewed, as the actual limits during power AMN Ansaldo Impianti, operation and constraints for reactor physics analyses (design and operation). Genoa, The basic elements of core management are also presented. The constraints on control rod movements during the achieving of criticality Italy and low power operation are illustrated in the fifth lecture. The banked position withdrawal sequence mode of operation is indeed an example of safety and design/operation combined approach. Some considerations on plant transient analyses are also presented in the fifth lecture, in order to show the impact Abstract between core and fuel performance and plant/system performance. The last (sixth) lecture is devoted to the open vessel testing during the These lectures on fundamentals of BWR reactor physics are a synthesis startup of a commercial BWR. A control .od calibration is also illustrated. of known and established concepts. These lecture are intended to be a comprehensive (even though descriptive in nature) presentation, which would give the basis for a fair understanding of power operation, fuel cycle and safety aspects of the boiling water reactor. The fundamentals of BWR reactor physics are oriented to design and operation.
    [Show full text]
  • A Graphite Isotope Ratio Method Primer—A Method for Estimating Plutonium Production in Graphite Moderated Reactors
    PNNL-14568 A Graphite Isotope Ratio Method Primer—A Method for Estimating Plutonium Production in Graphite Moderated Reactors C. J. Gesh February 2004 Prepared for the U.S. Department of Energy under Contract DE-AC06-76RL01830 DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor Battelle Memorial Institute, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or 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 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 Battelle Memorial Institute. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof. PACIFIC NORTHWEST NATIONAL LABORATORY operated by BATTELLE for the UNITED STATES DEPARTMENT OF ENERGY under Contract DE-AC06-76RL01830 This document was printed on recycled paper. PNNL-14568 A Graphite Isotope Ratio Method Primer—A Method for Estimating Plutonium Production in Graphite Moderated Reactors C. J. Gesh February 2004 Prepared for the U.S. Department of Energy under Contract DE-AC06-76RL01830 Pacific Northwest National Laboratory Richland, Washington 99352 Summary The Graphite Isotope Ratio Method (GIRM) is a technique used to estimate the total plutonium production in a graphite-moderated reactor.
    [Show full text]
  • NUCLEAR TECHNOLOGY REVIEW 2011 the Following States Are Members of the International Atomic Energy Agency
    Atoms for Peace Atoms for Peace 11-17411_NTR_cover.indd 1 2011-09-05 15:14:02 NUCLEAR TECHNOLOGY REVIEW 2011 The following States are Members of the International Atomic Energy Agency: AFGHANISTAN GHANA NORWAY ALBANIA GREECE OMAN ALGERIA GUATEMALA PAKISTAN ANGOLA HAITI PALAU ARGENTINA HOLY SEE PANAMA ARMENIA HONDURAS PARAGUAY AUSTRALIA HUNGARY PERU AUSTRIA ICELAND PHILIPPINES AZERBAIJAN INDIA POLAND BAHRAIN INDONESIA PORTUGAL BANGLADESH IRAN, ISLAMIC REPUBLIC OF QATAR BELARUS IRAQ REPUBLIC OF MOLDOVA BELGIUM IRELAND ROMANIA BELIZE ISRAEL RUSSIAN FEDERATION BENIN ITALY BOLIVIA JAMAICA SAUDI ARABIA BOSNIA AND HERZEGOVINA JAPAN SENEGAL BOTSWANA JORDAN SERBIA BRAZIL KAZAKHSTAN SEYCHELLES BULGARIA KENYA SIERRA LEONE BURKINA FASO KOREA, REPUBLIC OF SINGAPORE BURUNDI KUWAIT SLOVAKIA CAMBODIA KYRGYZSTAN SLOVENIA CAMEROON LATVIA SOUTH AFRICA CANADA LEBANON SPAIN CENTRAL AFRICAN LESOTHO SRI LANKA REPUBLIC LIBERIA SUDAN CHAD LIBYAN ARAB JAMAHIRIYA SWEDEN CHILE LIECHTENSTEIN SWITZERLAND CHINA LITHUANIA SYRIAN ARAB REPUBLIC COLOMBIA LUXEMBOURG TAJIKISTAN CONGO MADAGASCAR THAILAND COSTA RICA MALAWI THE FORMER YUGOSLAV CÔTE D’IVOIRE MALAYSIA REPUBLIC OF MACEDONIA CROATIA MALI TUNISIA CUBA MALTA TURKEY CYPRUS MARSHALL ISLANDS UGANDA CZECH REPUBLIC MAURITANIA UKRAINE DEMOCRATIC REPUBLIC MAURITIUS UNITED ARAB EMIRATES OF THE CONGO MEXICO DENMARK MONACO UNITED KINGDOM OF DOMINICAN REPUBLIC MONGOLIA GREAT BRITAIN AND ECUADOR MONTENEGRO NORTHERN IRELAND EGYPT MOROCCO UNITED REPUBLIC EL SALVADOR MOZAMBIQUE OF TANZANIA ERITREA MYANMAR UNITED STATES OF AMERICA ESTONIA NAMIBIA URUGUAY ETHIOPIA NEPAL UZBEKISTAN FINLAND NETHERLANDS VENEZUELA FRANCE NEW ZEALAND VIETNAM GABON NICARAGUA YEMEN GEORGIA NIGER ZAMBIA GERMANY NIGERIA ZIMBABWE The Agency’s Statute was approved on 23 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 1957.
    [Show full text]