General Chapter <1821> Radioactivity—Theory and Practice
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Radioisotopes and Radiopharmaceuticals
RADIOISOTOPES AND RADIOPHARMACEUTICALS Radioisotopes are the unstable form of an element that emits radiation to become a more stable form — they have certain special attributes. These make radioisotopes useful in areas such as medicine, where they are used to develop radiopharmaceuticals, as well as many other industrial applications. THE PRODUCTION OF TECHNETIUM-99m RADIOPHARMACEUTICALS: ONE POSSIBLE ROUTE IRRADIATED U-235 99 TARGETS MO PROCESSING FACILITY HOSPITAL RADIOPHARMACY (MIXING WITH BIOLOGICAL MOLECULES THAT BIND AT DIFFERENT LOCATIONS IN THE 99mTC IS IDEAL FOR DIAGNOSTICS BECAUSE OF BODY, SUPPORTING A WIDE ITS SHORT HALF-LIFE (6 HOURS) AND IDEAL GAMMA EMISSION RANGE OF MEDICAL NUCLEAR APPLICATIONS) REACTOR: 6 HOURS MAKES DISTRIBUTION DIFFICULT 99MO BULK LIQUID TARGET ITS PARENT NUCLIDE, MOLYBDENUM-99, IS PRODUCED; ITS HALF-LIFE (66 HOURS) MAKES 99 99m IRRADIATION IT SUITABLE FOR TRANSPORT MO/ TC GENERATORS 99MO/99mTC GENERATORS ARE PRODUCED AND DISTRIBUTED AROUND THE GLOBE 99MO/99TC GENERATOR MANUFACTURER Radioisotopes can occur naturally or be produced artificially, mainly in research reactors and accelerators. They are used in various fields, including nuclear medicine, where radiopharmaceuticals play a major role. Radiopharmaceuticals are substances that contain a radioisotope, and have properties that make them effective markers in medical diagnostic or therapeutic procedures. The chemical presence of radiopharmaceuticals can relay detailed information to medical professionals that can help in diagnoses and treatments. Eighty percent of all diagnostic medical scans worldwide use 99mTc, and its availability, at present, is dependent on the production of 99Mo in research reactors. Globally, the number of medical procedures involving the use of radioisotopes is growing, with an increasing emphasis on radionuclide therapy using radiopharmaceuticals for the treatment of cancer.. -
OPERATIONAL GUIDANCE on HOSPITAL RADIOPHARMACY: a SAFE and EFFECTIVE APPROACH the Following States Are Members of the International Atomic Energy Agency
OPERATIONAL GUIDANCE ON HOSPITAL RADIOPHARMACY: A SAFE AND EFFECTIVE APPROACH The following States are Members of the International Atomic Energy Agency: AFGHANISTAN GUATEMALA PAKISTAN ALBANIA HAITI PALAU ALGERIA HOLY SEE PANAMA ANGOLA HONDURAS PARAGUAY ARGENTINA HUNGARY PERU ARMENIA ICELAND PHILIPPINES AUSTRALIA INDIA POLAND AUSTRIA INDONESIA PORTUGAL AZERBAIJAN IRAN, ISLAMIC REPUBLIC OF QATAR BANGLADESH IRAQ REPUBLIC OF MOLDOVA BELARUS IRELAND ROMANIA BELGIUM ISRAEL RUSSIAN FEDERATION BELIZE ITALY SAUDI ARABIA BENIN JAMAICA SENEGAL BOLIVIA JAPAN SERBIA BOSNIA AND HERZEGOVINA JORDAN SEYCHELLES BOTSWANA KAZAKHSTAN BRAZIL KENYA SIERRA LEONE BULGARIA KOREA, REPUBLIC OF SINGAPORE BURKINA FASO KUWAIT SLOVAKIA CAMEROON KYRGYZSTAN SLOVENIA CANADA LATVIA SOUTH AFRICA CENTRAL AFRICAN LEBANON SPAIN REPUBLIC LIBERIA SRI LANKA CHAD LIBYAN ARAB JAMAHIRIYA SUDAN CHILE LIECHTENSTEIN SWEDEN CHINA LITHUANIA SWITZERLAND COLOMBIA LUXEMBOURG SYRIAN ARAB REPUBLIC COSTA RICA MADAGASCAR TAJIKISTAN CÔTE D’IVOIRE MALAWI THAILAND CROATIA MALAYSIA THE FORMER YUGOSLAV CUBA MALI REPUBLIC OF MACEDONIA CYPRUS MALTA TUNISIA CZECH REPUBLIC MARSHALL ISLANDS TURKEY DEMOCRATIC REPUBLIC MAURITANIA UGANDA OF THE CONGO MAURITIUS UKRAINE DENMARK MEXICO UNITED ARAB EMIRATES DOMINICAN REPUBLIC MONACO UNITED KINGDOM OF ECUADOR MONGOLIA GREAT BRITAIN AND EGYPT MONTENEGRO NORTHERN IRELAND EL SALVADOR MOROCCO ERITREA MOZAMBIQUE UNITED REPUBLIC ESTONIA MYANMAR OF TANZANIA ETHIOPIA NAMIBIA UNITED STATES OF AMERICA FINLAND NEPAL URUGUAY FRANCE NETHERLANDS UZBEKISTAN GABON NEW ZEALAND VENEZUELA GEORGIA NICARAGUA VIETNAM GERMANY NIGER YEMEN GHANA NIGERIA ZAMBIA GREECE NORWAY 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. The Headquarters of the Agency are situated in Vienna. -
Dead Time and Count Loss Determination for Radiation Detection Systems in High Count Rate Applications
Scholars' Mine Doctoral Dissertations Student Theses and Dissertations Spring 2010 Dead time and count loss determination for radiation detection systems in high count rate applications Amol Patil Follow this and additional works at: https://scholarsmine.mst.edu/doctoral_dissertations Part of the Nuclear Engineering Commons Department: Nuclear Engineering and Radiation Science Recommended Citation Patil, Amol, "Dead time and count loss determination for radiation detection systems in high count rate applications" (2010). Doctoral Dissertations. 2148. https://scholarsmine.mst.edu/doctoral_dissertations/2148 This thesis is brought to you by Scholars' Mine, a service of the Missouri S&T Library and Learning Resources. This work is protected by U. S. Copyright Law. Unauthorized use including reproduction for redistribution requires the permission of the copyright holder. For more information, please contact [email protected]. DEAD TIME AND COUNT LOSS DETERMINATION FOR RADIATION DETECTION SYSTEMS IN HIGH COUNT RATE APPLICATIONS by AMOL PATIL A DISSERTATION Presented to the Faculty of the Graduate School of the MISSOURI UNIVERSITY OF SCIENCE AND TECHNOLOGY In Partial Fulfillment of the Requirements for the Degree DOCTOR OF PHILOSOPHY in NUCLEAR ENGINEERING 2010 Approved by Shoaib Usman, Advisor Arvind Kumar Gary E. Mueller Carlos H. Castano Bijaya J. Shrestha © 2010 AMOL PATIL All Rights Reserved iii PUBLICATION DISSERTATION OPTION This dissertation consists of the following two articles that have been, or will be submitted for publication as follows: Pages 4-40 are intended for submission to Journal of Radioanalytical and Nuclear Chemistry. Pages 41-62 have been published in Nuclear Technologies journal (February, 2009). iv ABSTRACT This research is focused on dead time and the subsequent count loss estimation in radiation detection systems. -
1 11. Nuclear Chemistry 11.1 Stable and Unstable Nuclides Very Large
11. Nuclear Chemistry Chemical reactions occur as a result of loosing/gaining and sharing electrons in the valance shell which is far away from the atomic nucleus as we described in previous chapters in chemical bonding. In chemical reactions identity of the elements (atomic) and the makeup of the nuclei (mass due to protons and neutrons) is preserved which is reflected in the Law of Conservation of mass. This idea of atomic nucleus is always stable was shattered as Henri Becquerel discovered radioactivity in uranium compound where uranium nuclei changes or undergo nuclear reactions where nuclei of an element is transformed into nuclei of different element(s) while emitting ionization radiation. Marie Curie also began a study of radioactivity in a different form of uranium ore called pitchblende and she discovered the existence of two more highly radioactive new elements radium and polonium formed as the products during the decay of unstable nuclide of uranium-235. Curie measure that the radiation emanated was proportional to the amount (moles or number of nuclides) of radioactive element present, and she proposed that radiation was a property nucleus of an unstable atom. The area of chemistry that focuses on the nuclear changes is called nuclear chemistry. What changes in a nuclide result from the loss of each of the following? a) An alpha particle. b) A gamma ray. c) An electron. d) A neutron. e) A proton. Answer: a), c), d), e) 11.1 Stable and Unstable Nuclides There are stable and unstable radioactive nuclides. Unstable nuclides emit subatomic particles, with alpha −α, beta −β, gamma −γ, proton-p, neutrons-n being the most common. -
3 Gamma-Ray Detectors
3 Gamma-Ray Detectors Hastings A Smith,Jr., and Marcia Lucas S.1 INTRODUCTION In order for a gamma ray to be detected, it must interact with matteu that interaction must be recorded. Fortunately, the electromagnetic nature of gamma-ray photons allows them to interact strongly with the charged electrons in the atoms of all matter. The key process by which a gamma ray is detected is ionization, where it gives up part or all of its energy to an electron. The ionized electrons collide with other atoms and liberate many more electrons. The liberated charge is collected, either directly (as with a proportional counter or a solid-state semiconductor detector) or indirectly (as with a scintillation detector), in order to register the presence of the gamma ray and measure its energy. The final result is an electrical pulse whose voltage is proportional to the energy deposited in the detecting medhtm. In this chapter, we will present some general information on types of’ gamma-ray detectors that are used in nondestructive assay (NDA) of nuclear materials. The elec- tronic instrumentation associated with gamma-ray detection is discussed in Chapter 4. More in-depth treatments of the design and operation of gamma-ray detectors can be found in Refs. 1 and 2. 3.2 TYPES OF DETECTORS Many different detectors have been used to register the gamma ray and its eneqgy. In NDA, it is usually necessary to measure not only the amount of radiation emanating from a sample but also its energy spectrum. Thus, the detectors of most use in NDA applications are those whose signal outputs are proportional to the energy deposited by the gamma ray in the sensitive volume of the detector. -
1. Gamma-Ray Detectors for Nondestructive Analysis P
1. GAMMA-RAY DETECTORS FOR NONDESTRUCTIVE ANALYSIS P. A. Russo and D. T. Vo I. INTRODUCTION AND OVERVIEW Gamma rays are used for nondestructive quantitative analysis of nuclear material. Knowledge of both the energy of the gamma ray and its rate of emission from the unknown mass of nuclear material is required to interpret most measurements of nuclear material quantities. Therefore, detection of gamma rays for nondestructive analysis of nuclear materials requires both spectroscopy capability and knowledge of absolute specific detector response. Some techniques nondestructively quantify attributes other than nuclear material mass, but all rely on the ability to distinguish elements or isotopes and measure the relative or absolute yields of their corresponding radiation signatures. All require spectroscopy and most require high resolution. Therefore, detection of gamma rays for quantitative nondestructive analysis (NDA) of the mass or of other attributes of nuclear materials requires spectroscopy. A previous book on gamma-ray detectors for NDA1 provided generic descriptions of three detector categories: inorganic scintillation detectors, semiconductor detectors, and gas-filled detectors. This report described relevant detector properties, corresponding spectral characteristics, and guidelines for choosing detectors for NDA. The current report focuses on significant new advances in detector technology in these categories. Emphasis here is given to those detectors that have been developed at least to the stage of commercial prototypes. The type of NDA application – fixed installation in a count room, portable measurements, or fixed installation in a processing line or other active facility (storage, shipping/receiving, etc.) – influences the choice of an appropriate detector. Some prototype gamma-ray detection techniques applied to new NDA approaches may revolutionize how nuclear materials are quantified in the future. -
Appendix B: Recommended Procedures
Recommended Procedures Appendix B Appendix B: Recommended Procedures Appendix B provides recommended procedures for tasks frequently performed in the laboratory. These procedures outline acceptable methods for meeting radiation safety requirements. The procedures are generic in nature, allowing for the diversity of research facilities, on campus. Contamination Survey Procedures Surveys are performed to monitor for the presence of contamination. Minimum survey frequencies are specified on the radiation permit. The surveys should be sufficiently extensive to allow confidence that there is no contamination. Common places to check for contamination are: bench tops, tools and equipment, floors, telephones, floors, door handles and drawer pulls, and computer keyboards. Types of Contamination Removable contamination can be readily transferred from one surface to another. Removable contamination may present an internal and external hazard because it can be picked up on the skin and ingested. Fixed contamination cannot be readily removed and generally does not present a significant hazard unless the material comes loose or is present large enough amounts to be an external hazard. Types of Surveys There are two types of survey methods used: 1) a direct (or meter) survey, and 2) a wipe (or smear) survey. Direct surveys, using a Geiger-Mueller (GM) detector or scintillation probe, can identify gross contamination (total contamination consisting of both fixed and removable contamination) but will detect only certain isotopes. Wipe surveys, using “wipes” such as cotton swabs or filter papers counted on a liquid scintillation counter or gamma counter can identify removable contamination only but will detect most isotopes used at the U of I. Wipe surveys are the most versatile and sensitive method of detecting low-level removable contamination in the laboratory. -
DRC-2016-012921.Pdf
CLN-SRT-011 R1.0 Page 2 of 33 Introduction ................................................................................................................................ 3 Sources of Radiation .................................................................................................................. 3 Radiation: Particle vs Electromagnetic ....................................................................................... 5 Ionizing and non- ionizing radiation ............................................................................................ 8 Radioactive Decay ....................................................................................................................10 Interaction of Radiation with Matter ...........................................................................................15 Radiation Detection and Measurement .....................................................................................17 Biological Effects of Radiation ...................................................................................................18 Radiation quantities and units ...................................................................................................18 Biological effects .......................................................................................................................20 Effects of Radiation by Biological Organization .........................................................................20 Mechanisms of biological damage ............................................................................................21 -
Portable Front-End Readout System for Radiation Detection
Portable Front-End Readout System for Radiation Detection by Maris Tali THESIS for the degree of MASTER OF SCIENCE (Master in Electronics) Faculty of Mathematics and Natural Sciences University of Oslo June 2015 Det matematisk- naturvitenskapelige fakultet Universitetet i Oslo Portable Front-End Readout System for Radiation Detection Maris Tali Portable Front-End Readout System for Radiation Detection Maris Tali Portable Front-End Readout System for Radiation Detection Acknowledgements First of all, I would like to thank my advisor, Ketil Røed, for giving me the opportunity to write my master thesis on such an interesting and challenging subject as radiation instrumentation and measurement. I would also like to thank him for his insight and guidance during my master thesis and for the exciting opportunities during my studies to work in a real experimental environment. I would also like to thank the Electronic Laboratory of the University of Oslo, specif- ically Halvor Strøm, David Bang and Stein Nielsen who all gave me valuable advice and assistance and without whom this master thesis could not have been finished. Lastly, I would like to thank the co-designer of the system whom I worked together with during my master thesis, Eino J. Oltedal. It is customary to write that half of ones master thesis belongs to ones partner. However, this is actually the case in this instance so I guess 2/3 of my master thesis is yours and I definitely could not have done this without you. So, thank you very much! Oslo, June 2015 Maris Tali I Maris Tali Portable Front-End Readout System for Radiation Detection Contents Acknowledgments I Glossary VI List of Figures X List of Tables X Abstract XI 1 Introduction 1 1.1 Background and motivation . -
Chapter 12 Monographs of 99Mtc Pharmaceuticals 12
Chapter 12 Monographs of 99mTc Pharmaceuticals 12 12.1 99mTc-Pertechnetate I. Zolle and P.O. Bremer Chemical name Chemical structure Sodium pertechnetate Sodium pertechnetate 99mTc injection (fission) (Ph. Eur.) Technetium Tc 99m pertechnetate injection (USP) 99m ± Pertechnetate anion ( TcO4) 99mTc(VII)-Na-pertechnetate Physical characteristics Commercial products Ec=140.5 keV (IT) 99Mo/99mTc generator: T1/2 =6.02 h GE Healthcare Bristol-Myers Squibb Mallinckrodt/Tyco Preparation Sodium pertechnetate 99mTc is eluted from an approved 99Mo/99mTc generator with ster- ile, isotonic saline. Generator systems differ; therefore, elution should be performed ac- cording to the manual provided by the manufacturer. Aseptic conditions have to be maintained throughout the operation, keeping the elution needle sterile. The total eluted activity and volume are recorded at the time of elution. The resulting 99mTc ac- tivity concentration depends on the elution volume. Sodium pertechnetate 99mTc is a clear, colorless solution for intravenous injection. The pH value is 4.0±8.0 (Ph. Eur.). Description of Eluate 99mTc eluate is described in the European Pharmacopeia in two specific monographs de- pending on the method of preparation of the parent radionuclide 99Mo, which is generally isolated from fission products (Monograph 124) (Council of Europe 2005a), or produced by neutron activation of metallic 98Mo-oxide (Monograph 283) (Council of Europe 2005b). Sodium pertechnetate 99mTc injection solution satisfies the general requirements of parenteral preparations stated in the European Pharmacopeia (Council of Europe 2004). The specific activity of 99mTc-pertechnetate is not stated in the Pharmacopeia; however, it is recommended that the eluate is obtained from a generator that is eluted regularly, 174 12.1 99mTc-Pertechnetate every 24 h. -
Optimization of a Gas Flow Proportional Counter for Alpha Decay
Optimization of a gas flow proportional counter for alpha decay measurements Elena Ceballos Romero Master Thesis Institut f¨urKernphysik Mathematisch-Naturwissenschaftliche Fakult¨at Westf¨alische Wilhelms-Universit¨atM¨unster Prof. Dr. Alfons Khoukaz October 2012 III I am among those who think that science has great beauty. A scientist in his laboratory is not only a technician: he is also a child placed before natural phenomena which impress him like a fairy tale. We should not allow it to be believed that all scientific progress can be reduced to mechanisms, machines, gearings, even though such machinery has its own beauty. -Marie Curie A magdalena, por ponerme en este camino. A mis padres, por siempre acompa~narmeen ´el. IV V I certify that I have independently written this thesis and no other sources than the mentioned ones have been used. Referent: Prof. Dr. Alfons Khoukaz Correferent: Dr. Mar´ıaVilla Alfageme VI Contents 1. Introduction 1 2. Introduction to natural radiations 5 2.1. Radioactivity . .5 2.1.1. Decay laws . .5 2.1.2. Activity . .7 2.2. Decays . .7 2.2.1. Alpha decay . .7 2.2.2. Beta decay . .9 2.2.3. Gamma decay . 11 3. Theoretical background: Gas-filled detectors 13 3.1. General properties . 13 3.1.1. Number of ion pairs formed . 14 3.1.2. Behaviour of charged particles in gases . 14 3.1.3. Operational modes of gas detectors . 15 3.2. Proportional counters: gas multiplication effect . 17 3.3. Gas flow detectors . 18 4. Experimental set-up 21 4.1. Detector . 21 4.1.1. -
Background and Derivation of ANS-5.4 Standard Fission Product Release Model
NUREG/CR-7003 PNNL-18490 Background and Derivation of ANS-5.4 Standard Fission Product Release Model Office of Nuclear Regulatory Research AVAILABILITY OF REFERENCE MATERIALS IN NRC PUBLICATIONS NRC Reference Material Non-NRC Reference Material As of November 1999, you may electronically access Documents available from public and special technical NUREG-series publications and other NRC records at libraries include all open literature items, such as NRC’s Public Electronic Reading Room at books, journal articles, and transactions, Federal http://www.nrc.gov/reading-rm.html. Register notices, Federal and State legislation, and Publicly released records include, to name a few, congressional reports. Such documents as theses, NUREG-series publications; Federal Register notices; dissertations, foreign reports and translations, and applicant, licensee, and vendor documents and non-NRC conference proceedings may be purchased correspondence; NRC correspondence and internal from their sponsoring organization. memoranda; bulletins and information notices; inspection and investigative reports; licensee event reports; and Commission papers and their attachments. Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are NRC publications in the NUREG series, NRC maintained at— regulations, and Title 10, Energy, in the Code of The NRC Technical Library Federal Regulations may also be purchased from one Two White Flint North of these two sources. 11545 Rockville Pike 1. The Superintendent of Documents Rockville, MD 20852–2738 U.S. Government Printing Office Mail Stop SSOP Washington, DC 20402–0001 These standards are available in the library for Internet: bookstore.gpo.gov reference use by the public. Codes and standards are Telephone: 202-512-1800 usually copyrighted and may be purchased from the Fax: 202-512-2250 originating organization or, if they are American 2.