Investigation of a Plutonium 238 Skin Puncture Event
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1 RC-6 Internal Dosimetry the Science and Art of Internal Dose
RC-6 Internal Dosimetry The science and art of internal dose assessment H. Doerfel a, A. Andrasi b, M. Bailey c, V. Berkovski d, E. Blanchardon e, C.-M. Castellani f, R. Cruz-Suarez g, G. Etherington c, C. Hurtgen h, B. LeGuen i, I. Malatova j, J. Marsh c, J. Stather c aIDEA System GmbH, Am Burgweg 4, 76227 Karlsruhe, Germany, [email protected] bKFKI Atomic Energy Research Institute, Budapest, Hungary cRadiation Protection Division, Health Protection Agency, Chilton, Didcot, UK dRadiation Protection Institute, Kiev, Ukraine eInstitut de Radioprotection et de Sûreté Nucleaire, Fontenay-aux-Roses, France fENEA Radiation Protection Institute, Bologna, Italia gInternational Atomic Energy Agency, Vienna, Austria hSCK•CEN Belgian Nuclear Research Centre, Mol, Belgium iElectricite de France, Saint-Denis, France jNRPI, Prague, Czech Republic 1 Abstract Doses from intakes of radionuclides cannot be measured but must be assessed from monitoring, such as whole body counting or urinary excretion measurements. Such assessments require application of a biokinetic model and estimation of the exposure time, material properties, etc. Because of the variety of parameters involved, the results of such assessments may vary over a wide range, according to the skill and the experience of the assessor. The refresher course gives an overview over the state of the art of the determination of internal dose. The first part of the course deals with the measuring techniques for individual incorporation monitoring i.e (i) direct measurement of activity in the whole body or organs, (ii) measurement of activity excreted with urine and faeces, and (iii) measurement of activity in the breathing zone or at the workplace, respectively. -
Forschungszentrum Karlsruhe Third European Intercomparison Exercise
Forschungszentrum Karlsruhe Technik und Umwelt Wissenschaftliche Berichte FZKA 6457 Third European Intercomparison Exercise on Internal Dose Assessment Results of a Research Programme in the Framework of the EULEP/EURADOS Action Group “Derivation of Parameter Values for Application to the New Model of the Human Respiratory Tract for Occupational Exposure” 1997-1999 H. Doerfel, A. Andrasi 1), M. R. Bailey 2), A. Birchall 2), C.-M. Castellani 3), C. Hurtgen 4), N. Jarvis 2), L. Johansson 5), B. LeGuen 6), G. Tarroni 3) Hauptabteilung Sicherheit 1) AERI, Budapest, Hungary 2) NRPB, Chilton Didcot, United Kingdom 3) ENEA, Bologna, Italy 4) CEN/SCK, Mol, Belgium 5) University of Umea, Umea, Sweden 6) EDF-GDF, Saint Denis Cedex, France Forschungszentrum Karlsruhe GmbH, Karlsruhe 2000 Abstract The EULEP/EURADOS Action Group „Derivation of Parameter Values for Application to the New Model of the Human Respiratory Tract for Occupational Exposure“ has initiated a new intercomparison exercise on internal dose assessment. During the last few years the ICRP has developed a new generation of more realistic internal dosimetry models, including the Human Respiratory Tract Model (ICRP Publication 66) and recycling systemic models for actinides. These models have been used to calculate dose coefficients, which have been adopted in the revised EURATOM Directive. This recent intercomparison exercise gave special consideration to the effects of the new models and the choice of input parameters on the assessment of internal doses from monitoring results. It also took into account some aspects which have not been considered in previous exercises, such as air monitoring, natural radionuclides, exposure of the public, artificially created cases and artificially reduced information. -
The International Commission on Radiological Protection: Historical Overview
Topical report The International Commission on Radiological Protection: Historical overview The ICRP is revising its basic recommendations by Dr H. Smith Within a few weeks of Roentgen's discovery of gamma rays; 1.5 roentgen per working week for radia- X-rays, the potential of the technique for diagnosing tion, affecting only superficial tissues; and 0.03 roentgen fractures became apparent, but acute adverse effects per working week for neutrons. (such as hair loss, erythema, and dermatitis) made hospital personnel aware of the need to avoid over- Recommendations in the 1950s exposure. Similar undesirable acute effects were By then, it was accepted that the roentgen was reported shortly after the discovery of radium and its inappropriate as a measure of exposure. In 1953, the medical applications. Notwithstanding these observa- ICRU recommended that limits of exposure should be tions, protection of staff exposed to X-rays and gamma based on consideration of the energy absorbed in tissues rays from radium was poorly co-ordinated. and introduced the rad (radiation absorbed dose) as a The British X-ray and Radium Protection Committee unit of absorbed dose (that is, energy imparted by radia- and the American Roentgen Ray Society proposed tion to a unit mass of tissue). In 1954, the ICRP general radiation protection recommendations in the introduced the rem (roentgen equivalent man) as a unit early 1920s. In 1925, at the First International Congress of absorbed dose weighted for the way different types of of Radiology, the need for quantifying exposure was radiation distribute energy in tissue (called the dose recognized. As a result, in 1928 the roentgen was equivalent in 1966). -
Medical Management of Persons Internally Contaminated with Radionuclides in a Nuclear Or Radiological Emergency
CONTAMINATION EPR-INTERNAL AND RESPONSE PREPAREDNESS EMERGENCY EPR-INTERNAL EPR-INTERNAL CONTAMINATION CONTAMINATION 2018 2018 2018 Medical Management of Persons Internally Contaminated with Radionuclides in a Nuclear or Radiological Emergency Contaminatedin aNuclear with Radionuclides Internally of Persons Management Medical Medical Management of Persons Internally Contaminated with Radionuclides in a Nuclear or Radiological Emergency A Manual for Medical Personnel Jointly sponsored by the Endorsed by AMERICAN SOCIETY FOR RADIATION ONCOLOGY INTERNATIONAL ATOMIC ENERGY AGENCY V I E N N A ISSN 2518–685X @ IAEA SAFETY STANDARDS AND RELATED PUBLICATIONS IAEA SAFETY STANDARDS Under the terms of Article III of its Statute, the IAEA is authorized to establish or adopt standards of safety for protection of health and minimization of danger to life and property, and to provide for the application of these standards. The publications by means of which the IAEA establishes standards are issued in the IAEA Safety Standards Series. This series covers nuclear safety, radiation safety, transport safety and waste safety. The publication categories in the series are Safety Fundamentals, Safety Requirements and Safety Guides. Information on the IAEA’s safety standards programme is available on the IAEA Internet site http://www-ns.iaea.org/standards/ The site provides the texts in English of published and draft safety standards. The texts of safety standards issued in Arabic, Chinese, French, Russian and Spanish, the IAEA Safety Glossary and a status report for safety standards under development are also available. For further information, please contact the IAEA at: Vienna International Centre, PO Box 100, 1400 Vienna, Austria. All users of IAEA safety standards are invited to inform the IAEA of experience in their use (e.g. -
Radiation Quantities and Units Definitions
Radiation Quantities and Units Definitions The quantities and units of radiation dose are inherently more complex than those used in toxicology or pharmacology, and additional complexity has resulted from several changes required by evolving concepts in radiation dosimetry. The first widely used physical quantity of radiation was “exposure,” related to the ability of x or gamma radiation to ionize air; its unit was the roentgen (R). Exposure was limited to photon radiation with energy less than 2.5 MeV. The quantity “absorbed dose” (D) was introduced because it was applicable to all forms of ionizing radiation and absorbing materials. Absorbed dose is energy deposited per unit mass, and its original unit was the rep (roentgen equivalent-physical); 1 rep equaled 93 ergs per g (0.0093 J per kg) of absorbing material. The rep was replaced with the rad (radiation absorbed dose); 1 rad equaled 100 ergs per g (0.01 J per kg). The “dose equivalent” (H) and its unit, the rem (roentgen equivalent-man), were introduced to account for the different biologic effects of the same absorbed dose from different types of radiation; H is the product of D, Q, and N at a point of interest in tissue, where D is absorbed dose, Q is the quality factor, and N is the product of any other modifying factors. The “effective dose equivalent” was introduced to include the different sensitivities of individual tissues and organs, which are important for internal dosimetry: its unit is the same as the unit of “dose equivalent.” In the 1990 recommendations of the International Commission on Radiological Protection (ICRP 1991), the use of N was dropped and the radiation weighting factor (WR) was substituted for Q. -
MIRD Pamphlet No. 22 - Radiobiology and Dosimetry of Alpha- Particle Emitters for Targeted Radionuclide Therapy
Alpha-Particle Emitter Dosimetry MIRD Pamphlet No. 22 - Radiobiology and Dosimetry of Alpha- Particle Emitters for Targeted Radionuclide Therapy George Sgouros1, John C. Roeske2, Michael R. McDevitt3, Stig Palm4, Barry J. Allen5, Darrell R. Fisher6, A. Bertrand Brill7, Hong Song1, Roger W. Howell8, Gamal Akabani9 1Radiology and Radiological Science, Johns Hopkins University, Baltimore MD 2Radiation Oncology, Loyola University Medical Center, Maywood IL 3Medicine and Radiology, Memorial Sloan-Kettering Cancer Center, New York NY 4International Atomic Energy Agency, Vienna, Austria 5Centre for Experimental Radiation Oncology, St. George Cancer Centre, Kagarah, Australia 6Radioisotopes Program, Pacific Northwest National Laboratory, Richland WA 7Department of Radiology, Vanderbilt University, Nashville TN 8Division of Radiation Research, Department of Radiology, New Jersey Medical School, University of Medicine and Dentistry of New Jersey, Newark NJ 9Food and Drug Administration, Rockville MD In collaboration with the SNM MIRD Committee: Wesley E. Bolch, A Bertrand Brill, Darrell R. Fisher, Roger W. Howell, Ruby F. Meredith, George Sgouros (Chairman), Barry W. Wessels, Pat B. Zanzonico Correspondence and reprint requests to: George Sgouros, Ph.D. Department of Radiology and Radiological Science CRB II 4M61 / 1550 Orleans St Johns Hopkins University, School of Medicine Baltimore MD 21231 410 614 0116 (voice); 413 487-3753 (FAX) [email protected] (e-mail) - 1 - Alpha-Particle Emitter Dosimetry INDEX A B S T R A C T......................................................................................................................... -
Perspective on Internal Contamination and Dose
Perspective on Internal Contamination and Dose Hanford Internal Dosimetry Program 1 What is meant by contamination and dose? Contamination: the unwanted presence of radioactive material. Dose: energy deposited in the body by radioactive decay of contamination 2 External and Internal Sources External exposure: Route of intake: Route of intake: no intake inhalation skin absorption Route of intake: Route of intake: wound exposure ingestion 3 Measuring External Dose Dosimeters measure dose from external sources. Left: Hanford Left: Hanford combination standard neutron dosimeter dosimeter Below: Hanford extremity dosimeter Dose from internal contamination (within the body) cannot be measured by dosimeters. 4 External Dose vs. Internal Dose External Dose Internal Dose • Generally, easy to measure with • Requires determination from a dosimeter bioassay measurements of • Usually reported as a “whole material in the body or excreted body” or “effective” dose from the body • Type of bioassay depends on radioactive material • Optimum type of bioassay depends on radioactive material and time since intake occurred 5 Bioassay Methods There are several bioassay methods used to determine intake and internal dose: • Whole body exam/count Left: a whole body • Chest count count and lung count being taken • Wound count • Urine/fecal sample analysis Left: the in-home kit for collecting urine or fecal samples for bioassay 6 How Critical is Time to Bioassay? Don’t sample too soon… but don’t wait too long • The material may not be present in the bladder or bowel if the sample is taken too soon. • Dose detection capability varies with time after intake. Getting a bioassay at the right time can provide a more accurate analysis. -
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. -
External and Internal Dosimetry
Chapter 7 External and Internal Dosimetry H-117 – Introductory Health Physics Slide 1 Objectives ¾ Discuss factors influencing external and internal doses ¾ Define terms used in external dosimetry ¾ Discuss external dosimeters such as TLDs, film badges, OSL dosimeters, pocket chambers, and electronic dosimetry H-117 – Introductory Health Physics Slide 2 Objectives ¾ Define terms used in internal dosimetry ¾ Discuss dose units and limits ¾ Define the ALI, DAC and DAC-hr ¾ Discuss radiation signs and postings H-117 – Introductory Health Physics Slide 3 Objectives ¾ Discuss types of bioassays ¾ Describe internal dose measuring equipment and facilities ¾ Discuss principles of internal dose calculation and work sample problems H-117 – Introductory Health Physics Slide 4 External Dosimetry H-117 – Introductory Health Physics Slide 5 External Dosimetry Gamma, beta or neutron radiation emitted by radioactive material outside the body irradiates the skin, lens of the eye, extremities & the whole body (i.e. internal organs) H-117 – Introductory Health Physics Slide 6 External Dosimetry (cont.) ¾ Alpha particles cannot penetrate the dead layer of skin (0.007 cm) ¾ Beta particles are primarily a skin hazard. However, energetic betas can penetrate the lens of an eye (0.3 cm) and deeper tissue (1 cm) ¾ Beta sources can produce more penetrating radiation through bremsstrahlung ¾ Primary sources of external exposure are photons and neutrons ¾ External dose must be measured by means of appropriate dosimeters H-117 – Introductory Health Physics Slide 7 -
Internal Dosimetry Co-Exposure Data for the Savannah River Site
Page 1 of 287 DOE Review Release 09/04/2020 ORAUT-OTIB-0081 Rev. 05 Internal Dosimetry Co-Exposure Data for the Effective Date: 09/01/2020 Savannah River Site Supersedes: Revision 04 Subject Expert(s): Matthew G. Arno, James M. Mahathy, and Nancy Chalmers Document Owner Signature on File Approval Date: 08/24/2020 Approval: Matthew G. Arno, Document Owner Concurrence: Signature on File Concurrence Date: 08/24/2020 John M. Byrne, Objective 1 Manager Concurrence: Signature on File Concurrence Date: 08/24/2020 Scott R. Siebert, Objective 3 Manager Concurrence: Vickie S. Short Signature on File for Concurrence Date: 08/24/2020 Kate Kimpan, Project Director Approval: Signature on File Approval Date: 08/25/2020 Timothy D. Taulbee, Associate Director for Science FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION. New Total Rewrite Revision Page Change Document No. ORAUT-OTIB-0081 Revision No. 05 Effective Date: 09/01/2020 Page 2 of 287 PUBLICATION RECORD EFFECTIVE REVISION DATE NUMBER DESCRIPTION 02/08/2013 00 New technical information bulletin to provide internal coworker data for the Savannah River Site. Incorporates formal internal and NIOSH review comments. Training required: As determined by the Objective Manager. Initiated by Matthew Arno. 04/01/2013 01 Revision initiated to correct the values provided in Tables 5-6, Type S uranium intake rates for 1968 through 2007, 5-10, changed end date from 2006 to 2007, A-3, plutonium bioassay data for 1955 through 2007, and A-8, neptunium bioassay data for 1991 through 2007. -
11. Dosimetry Fundamentals
Outline • Introduction Dosimetry Fundamentals • Dosimeter model • Interpretation of dosimeter measurements Chapter 11 – Photons and neutrons – Charged particles • General characteristics of dosimeters F.A. Attix, Introduction to Radiological Physics and Radiation Dosimetry • Summary Introduction Dosimeter • Radiation dosimetry deals with the determination • A dosimeter can be generally defined as (i.e., by measurement or calculation) of the any device that is capable of providing a absorbed dose or dose rate resulting from the interaction of ionizing radiation with matter reading R that is a measure of the absorbed • Other radiologically relevant quantities are dose Dg deposited in its sensitive volume V exposure, kerma, fluence, dose equivalent, energy by ionizing radiation imparted, etc. can be determined • If the dose is not homogeneous • Measuring one quantity (usually the absorbed dose) another one can be derived through throughout the sensitive calculations based on the defined relationships volume, then R is a measure of mean value Dg Dosimeter Simple dosimeter model • Ordinarily one is not interested in measuring • A dosimeter can generally be considered as the absorbed dose in a dosimeter’s sensitive consisting of a sensitive volume V filled with a volume itself, but rather as a means of medium g, surrounded by a wall (or envelope, determining the dose (or a related quantity) for container, etc.) of another medium w having a another medium in which direct measurements thickness t 0 are not feasible • A simple dosimeter can be -
Internal Dosimetry Hazard and Risk Assessments: Methods and Applications G
DOSIMETRY AND INSTRUMENTATION INTERNAL DOSIMETRY HAZARD AND RISK ASSESSMENTS: METHODS AND APPLICATIONS G. A. Roberts RWE NUKEM Limited, Didcot, UK ABSTRACT Routine internal dose exposures are typically (in the UK nuclear industry) less than external dose exposures: however, the costs of internal dosimetry monitoring programmes can be significantly greater than those for external dosimetry. For this reason decisions on when to apply routine monitoring programmes, and the nature of these programmes, can be more critical than for external dosimetry programmes. This paper describes various methods for performing hazard and risk assessments which are being developed by RWE NUKEM Limited Approved Dosimetry Services to provide an indication when routine internal dosimetry monitoring should be considered. INTRODUCTION The estimation of internal dose needs to be inferred from the measurement of activity in the individual (by in-vivo or in-vitro methods) or in the workplace environment. A variety of monitoring methods are available for making these measurements [1]. A review of the current monitoring practices within European Union countries was included within the OMINEX project (Optimisation of Monitoring for Internal Exposure) [2] and are summarised in Table 1. Type of Total Monitoring method Monitoring operation number In vivo Bioassay SAS/PAS Routine Special Both Nuclear power 34 33 15 7 8 4 27 plant Reprocessing 2 2 2 - - - 1 Fuel 7 6 6 2 3 2 5 fabrication Decommiss- 6 5 6 1 2 2 5 ioning Research 19 17 12 5 3 5 16 Non-nuclear 3 3 3 2 2 2 2 SAS – static air sampling PAS – personal air sampling Table 1: Number of organisations using each monitoring method for routine or special monitoring, classified by type of operation.