Various Measuring Instruments
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Nuclear Threat
MANAGEMENT OF RADIOLOGIC CASUALTIES Nuclear Threat . Formerly .Soviet Union .Nuclear fallout . Now .NBC threat against civilians .Accidental exposure of workers and public Images: CIA, FBI Accidental Exposure in Brazil . Cesium-137 source found by scavengers in 1987 . Source broken open, contents shared . 112,800 surveyed for contamination . 120 externally contaminated only . 129 internally & externally contaminated . 20 required hospital treatment . 14 developed bone marrow depression . 8 treated with Granulocyte Macrophage Colony-Stimulating Factors (GM-CSF) . 4 died acute phase, hemorrhage, infection . 1 died in 1994 from liver failure Images: CIA Medical Staff Exposure Medical staff received doses: . Maximum 500 millirem (5 millisieverts) .Natural background radiation dose ~ 200 millirems annually . Average 20 millirem (0.2 millisieverts) .Equivalent to one chest X-ray Juarez, Mexico Incident . 400 curies of cobalt-60 in stainless steel therapy device sold for scrap . Ended up in recycled steel rebar . Wrong turn into Los Alamos lab . I0 people significantly exposed . 1 construction worker died .bone cancer . 109 houses demolished in Mexico Images: CIA US Experience 1944-1999 . 243 radiation accidents leading to “serious” classification . 790 people received significant exposure resulting in 30 fatalities .Incidents included: . 137 industrial . 80 medical . 11 criticality Image: DOE The Basics of Radiation Radiation is energy that comes from a source and travels through matter or space Radioactivity is the spontaneous emission of radiation: . Either directly from unstable atomic nuclei, or . As a consequence of a nuclear reaction, or . Machine – produced (X-ray) Image: NOAA Contamination . Defined as internal or external deposition of radioactive particles . Irradiation continues until source removed by washing, flushing or radioactive decay . -
Personal Radiation Monitoring
Personal Radiation Monitoring Tim Finney 2020 Radiation monitoring Curtin staff and students who work with x-ray machines, neutron generators, or radioactive substances are monitored for exposure to ionising radiation. The objective of radiation monitoring is to ensure that existing safety procedures keep radiation exposure As Low As Reasonably Achievable (ALARA). Personal radiation monitoring badges Radiation exposure is measured using personal radiation monitoring badges. Badges contain a substance that registers how much radiation has been received. Here is the process by which a user’s radiation dose is measured: 1. The user is given a badge to wear 2. The user wears the badge for a set time period (usually three months) 3. At the end of the set time, the user returns the badge 4. The badge is sent away to be read 5. A dose report is issued. These steps are repeated until monitoring is no longer required. Badges are supplied by a personal radiation monitoring service provider. Curtin uses a service provider named Landauer. In addition to user badges, the service provider sends control badges that are kept on site in a safe place away from radiation sources. The service provider reads each badge using a process that extracts a signal from the substance contained in the badge to obtain a dose measurement. (Optically stimulated luminescence is one such process.) The dose received by the control badge is subtracted from the user badge reading to obtain the user dose during the monitoring period. Version 1.0 Uncontrolled document when printed Health and Safety Page 1 of 7 A personal radiation monitoring badge Important Radiation monitoring badges do not protect you from radiation exposure. -
Radiation Survey Meters in Hospitals: Technology, Challenges, and a New Approach
White paper Radiation survey meters in hospitals: technology, challenges, and a new approach This white paper introduces a new radiation sur- essary environmental radiation in hospitals. To vey meter for the hospital segment: the RaySafe lower the risks of unwanted radiation-induced 452. First, the need for survey meters in hospi- effects, the goal is to minimize the exposure to tal environments is described. This is followed unnecessary radiation, both for patients and staff. by an introduction to common survey meter To be able to take the correct measures, it is es- technologies to illustrate their possibilities and sential to know the properties of the radiation in delimitations. Next, some difficulties with survey terms of energy range, dose rate, type of radia- meter measurements in hospitals are identified tion, and more. This is where radiation survey and discussed. Finally, the technology inside the meters are needed. RaySafe 452 is presented to summarize how the instrument addresses the identified issues. Figure 1 illustrates some measurement scenarios and demonstrate the wide range of measurement 1. Why radiation survey meters are need- needs in hospitals. Table 1 gives an overview of ed in hospitals the main areas of application of survey meters in hospitals, with quantities, units, and typical Ionizing radiation is not visible to the human eye. energy ranges of the radiation. It does not smell or make a noise, and you cannot feel it. In the complex and busy environment of a hospital, how can you ensure that patients and staff are not exposed to more ionizing radiation than necessary? Ionizing radiation is utilized for different purposes in medical procedures, with the three main areas being diagnostic imaging, nuclear medicine, and radiotherapy. -
Proper Use of Radiation Dose Metric Tracking for Patients Undergoing Medical Imaging Exams
Proper Use of Radiation Dose Metric Tracking for Patients Undergoing Medical Imaging Exams Frequently Asked Questions Introduction In August of 2021, the American Association of Physicists in Medicine (AAPM), the American College of Radiology (ACR), and the Health Physics Society (HPS) jointly released the following position statement advising against using information about a patient’s previous cumulative dose information from medical imaging exams to decide the appropriateness of future imaging exams. This statement was also endorsed by the Radiological Society of North America (RSNA). It is the position of the American Association of Physicists in Medicine (AAPM), the American College of Radiology (ACR), and the Health Physics Society (HPS) that the decision to perform a medical imaging exam should be based on clinical grounds, including the information available from prior imaging results, and not on the dose from prior imaging-related radiation exposures. AAPM has long advised, as recommended by the International Commission on Radiological Protection (ICRP), that justification of potential patient benefit and subsequent optimization of medical imaging exposures are the most appropriate actions to take to protect patients from unnecessary medical exposures. This is consistent with the foundational principles of radiation protection in medicine, namely that patient radiation dose limits are inappropriate for medical imaging exposures. Therefore, the AAPM recommends against using dose values, including effective dose, from a patient’s prior imaging exams for the purposes of medical decision making. Using quantities such as cumulative effective dose may, unintentionally or by institutional or regulatory policy, negatively impact medical decisions and patient care. This position statement applies to the use of metrics to longitudinally track a patient’s dose from medical radiation exposures and infer potential stochastic risk from them. -
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 . -
Cumulative Radiation Dose in Patients Admitted with Subarachnoid Hemorrhage: a Prospective PATIENT SAFETY Study Using a Self-Developing Film Badge
Cumulative Radiation Dose in Patients Admitted with Subarachnoid Hemorrhage: A Prospective PATIENT SAFETY Study Using a Self-Developing Film Badge A.C. Mamourian BACKGROUND AND PURPOSE: While considerable attention has been directed to reducing the x-ray H. Young dose of individual imaging studies, there is little information available on the cumulative dose during imaging-intensive hospitalizations. We used a radiation-sensitive badge on 12 patients admitted with M.F. Stiefel SAH to determine if this approach was feasible and to measure the extent of their x-ray exposure. MATERIALS AND METHODS: After obtaining informed consent, we assigned a badge to each of 12 patients and used it for all brain imaging studies during their ICU stay. Cumulative dose was deter- mined by quantifying exposure on the badge and correlating it with the number and type of examinations. RESULTS: The average skin dose for the 3 patients who had only diagnostic DSA without endovascular intervention was 0.4 Gy (0.2–0.6 Gy). The average skin dose of the 8 patients who had both diagnostic DSA and interventions (eg, intra-arterial treatment of vasospasm and coiling of aneurysms) was 0.9 Gy (1.8–0.4 Gy). One patient had only CT examinations. There was no effort made to include or exclude the badge in the working view during interventions. CONCLUSIONS: It is feasible to incorporate a film badge that uses a visual scale to monitor the x-ray dose into the care of hospitalized patients. Cumulative skin doses in excess of 1 Gy were not uncommon (3/12) in this group of patients with acute SAH. -
Radiation Glossary
Radiation Glossary Activity The rate of disintegration (transformation) or decay of radioactive material. The units of activity are Curie (Ci) and the Becquerel (Bq). Agreement State Any state with which the U.S. Nuclear Regulatory Commission has entered into an effective agreement under subsection 274b. of the Atomic Energy Act of 1954, as amended. Under the agreement, the state regulates the use of by-product, source, and small quantities of special nuclear material within said state. Airborne Radioactive Material Radioactive material dispersed in the air in the form of dusts, fumes, particulates, mists, vapors, or gases. ALARA Acronym for "As Low As Reasonably Achievable". Making every reasonable effort to maintain exposures to ionizing radiation as far below the dose limits as practical, consistent with the purpose for which the licensed activity is undertaken. It takes into account the state of technology, the economics of improvements in relation to state of technology, the economics of improvements in relation to benefits to the public health and safety, societal and socioeconomic considerations, and in relation to utilization of radioactive materials and licensed materials in the public interest. Alpha Particle A positively charged particle ejected spontaneously from the nuclei of some radioactive elements. It is identical to a helium nucleus, with a mass number of 4 and a charge of +2. Annual Limit on Intake (ALI) Annual intake of a given radionuclide by "Reference Man" which would result in either a committed effective dose equivalent of 5 rems or a committed dose equivalent of 50 rems to an organ or tissue. Attenuation The process by which radiation is reduced in intensity when passing through some material. -
``Low Dose``And/Or``High Dose``In Radiation Protection: a Need To
IAEA-CN-67/154 "LOW DOSE" AND/OR "HIGH DOSE" IN RADIATION PROTECTION: A NEED TO SETTING CRITERIA FOR DOSE CLASSIFICATION Mehdi Sohrabi National Radiation Protection Department & XA9745670 Center for Research on Elevated Natural Radiation Atomic Energy Organization of Iran, Tehran ABSTRACT The "low dose" and/or "high dose" of ionizing radiation are common terms widely used in radiation applications, radiation protection and radiobiology, and natural radiation environment. Reading the title, the papers of this interesting and highly important conference and the related literature, one can simply raise the question; "What are the levels and/or criteria for defining a low dose or a high dose of ionizing radiation?". This is due to the fact that the criteria for these terms and for dose levels between these two extreme quantities have not yet been set, so that the terms relatively lower doses or higher doses are usually applied. Therefore, setting criteria for classification of radiation doses in the above mentioned areas seems a vital need. The author while realizing the existing problems to achieve this important task, has made efforts in this paper to justify this need and has proposed some criteria, in particular for the classification of natural radiation areas, based on a system of dose limitation. In radiation applications, a "low dose" is commonly associated to applications delivering a dose such as given to a patient in a medical diagnosis and a "high dose" is referred to a dose in applications such as radiotherapy, mutation breeding, control of sprouting, control of insects, delay ripening, and sterilization having a range of doses from 1 to 105 Gy, for which the term "high dose dosimetry" is applied [1].