DOCSLIB.ORG

Radiation Protection Qualification Standard Reference Guide MARCH 2009

Total Page：16

File Type：pdf, Size：1020Kb

Download full-text PDF Read full-text
Radiation Protection Qualification Standard Reference Guide MARCH 2009 Radiation Protection Qualification Standard Reference Guide MARCH 2009 This page intentionally left blank ii Table of Contents LIST OF FIGURES ..................................................................................................................... iv LIST OF TABLES ....................................................................................................................... iv ACRONYMS, ABBREVIATIONS and SYMBOLS ................................................................. v PURPOSE...................................................................................................................................... 1 SCOPE ........................................................................................................................................... 1 PREFACE...................................................................................................................................... 1 TECHNICAL COMPETENCIES............................................................................................... 3 1. Radiation protection personnel shall demonstrate a working level knowledge of the various types of radiation and how they interact with matter................................................... 3 2. Radiation protection personnel shall demonstrate a working level knowledge of radioactivity and transformation mechanisms. ....................................................................... 19 3. Radiation protection personnel shall demonstrate a working level knowledge of principles and concepts for internal and external dosimetry. ................................................. 25 4. Radiation protection personnel shall demonstrate a working level knowledge of the biological effects of radiation. ................................................................................................ 42 5. Radiation Protection personnel shall demonstrate a working level knowledge of the principles and use of radiological instrumentation and radiological monitoring/survey practices. ................................................................................................................................. 48 6. Radiation protection personnel shall demonstrate a working level knowledge of internal and external radiation protection principles and control techniques. ..................................... 68 7. Radiation protection personnel shall demonstrate a working level knowledge of ALARA principles, and their application to radiological work activities............................................. 77 8. Radiation protection personnel shall demonstrate a working level knowledge of the application of engineered radiological controls and facility design, including containment/confinement systems.......................................................................................... 84 9. Radiation protection personnel shall demonstrate a familiarity level knowledge of the radiological hazards associated with the following and a working level knowledge for site specific radiological hazards: ........................................................................................... 91 10. Radiation protection personnel shall demonstrate an expert level knowledge of the DOE radiation protection system for occupational workers as set forth in the following policy, requirements, and guidance documents:............................................................................... 110 11. Radiation protection personnel shall demonstrate a working level knowledge of the following DOE Policy, Order, and Manual Directives, and Technical Standards related to radiation protection:.......................................................................................................... 138 12. Radiation protection personnel shall demonstrate a familiarity level knowledge of the identification, reporting, investigation, and enforcement related to potential noncompliance with nuclear safety requirements................................................................. 143 13. Radiation protection personnel shall demonstrate a familiarity level knowledge of radioactive waste management: ............................................................................................ 146 14. Radiation protection personnel shall demonstrate a working level knowledge of DOE requirements and guidance related to safety management. .................................................. 150 15. Radiation protection personnel shall demonstrate a familiarity level knowledge of Federal regulations and DOE Orders related to emergency planning and preparedness as they pertain to radiological incidents.................................................................................... 161 16. Radiation protection personnel shall demonstrate a familiarity level knowledge of DOE Orders related to Federal and contractor personnel training and qualification..................... 168 iii 17. Radiation protection personnel shall demonstrate a working level knowledge of national and international radiation protection standards and recommendations............................... 170 18. Radiation protection personnel shall demonstrate a familiarity level knowledge of the Federal regulations, guidelines, and DOE Orders pertaining to the decontamination and decommissioning of nuclear facilities. ................................................................................. 175 19. Radiation protection personnel shall demonstrate a familiarity level knowledge of the standards and DOE Orders pertaining to the packaging and transportation of radioactive materials................................................................................................................................ 188 20. Radiation protection personnel shall demonstrate a familiarity level knowledge of the Department’s philosophy and approach to implementing Integrated Safety Management.. 189 21. Radiation protection personnel shall demonstrate a familiarity level knowledge of the Department’s guidance for the structure, function, and operation of a radiation generating device (RGD) control program as discussed in Implementation Guide G 441.1-5, Radiation Generating Devices. ........................................................................... 195 Selected Bibliography and Suggested Reading ...................................................................... 198 LIST OF FIGURES Figure 1. Ionization by an alpha particle.........................................................................................6 Figure 2. Bremsstrahlung radiation.................................................................................................8 Figure 3. Elastic scattering..............................................................................................................9 Figure 4. Inelastic scattering ...........................................................................................................9 Figure 5. Photoelectric effect........................................................................................................11 Figure 6. Compton scattering........................................................................................................11 Figure 7. Pair production and annihilation....................................................................................12 Figure 8. Electron entrapment.......................................................................................................33 Figure 9. Energy dependence of various TLD-Albedo dosimeters ..............................................34 Figure 10. Voltage-response curve ...............................................................................................51 Figure 11. Conductor ....................................................................................................................53 Figure 12. Insulator.......................................................................................................................53 Figure 13. Semiconductor.............................................................................................................53 Figure 14. Electron hole................................................................................................................54 Figure 15. n-p junction..................................................................................................................55 Figure 16. High-resolution gamma ray spectroscopy system.......................................................57 Figure 17. High-volume sampler ..................................................................................................61 Figure 18. Low-volume sampler...................................................................................................62 LIST OF TABLES Table 1. Density-thickness values for human body ......................................................................37 Table 2. Biological effects of short term radiation on humans .....................................................45 Table 3. Suggested radiological control performance indicators..................................................81 Table 4. Radioactive decay properties of selected isotopes and decay products, excluding spontaneous fission ......................................................................................................94 Table 5. Half-lives and natural percent abundance for important uranium isotopes in the nuclear fuel cycle .........................................................................................................98
Load more

Recommended publications
  • Chapter 1.2: Transformation Kinetics
    Chapter 1: Radioactivity Chapter 1.2: Transformation Kinetics 200 NPRE 441, Principles of Radiation Protection Chapter 1: Radioactivity http://www.world‐nuclear.org/info/inf30.html 201 NPRE 441, Principles of Radiation Protection Chapter 1: Radioactivity Serial Transformation In many situations, the parent nuclides produce one or more radioactive offsprings in a chain. In such cases, it is important to consider the radioactivity from both the parent and the daughter nuclides as a function of time. • Due to their short half lives, 90Kr and 90Rb will be completely transformed, results in a rapid building up of 90Sr. • 90Y has a much shorter half‐life compared to 90Sr. After a certain period of time, the instantaneous amount of 90Sr transformed per unit time will be equal to that of 90Y. •Inthiscase,90Y is said to be in a secular equilibrium. 202 NPRE 441, Principles of Radiation Protection Chapter 1: Radioactivity Exponential DecayTransformation Kinetics • Different isotopes are characterized by their different rate of transformation (decay). • The activity of a pure radionuclide decreases exponentially with time. For a given sample, the number of decays within a unit time window around a given time t is a Poisson random variable, whose expectation is given by t Q Q0e •Thedecayconstant is the probability of a nucleus of the isotope undergoing a decay within a unit period of time. 203 NPRE 441, Principles of Radiation Protection Chapter 1: Radioactivity Why Exponential Decay? The decay rate, A, is given by Separate variables in above equation, we have The decay constant is the probability of a nucleus of the isotope undergoing a decay within a unit period of time.
    [Show full text]
  • Glossary Derived From: Human Research Program Integrated Research Plan, Revision A, (January 2009)
    Glossary derived from: Human Research Program Integrated Research Plan, Revision A, (January 2009). National Aeronautics and Space Administration, Johnson Space Center, Houston, Texas 77058, pages 232-280. Report No. 153: Information Needed to Make Radiation Protection Recommendations for Space Missions Beyond Low-Earth Orbit (2006). National Council on Radiation Protection and Measurements, pages 309-318. Reprinted with permission of the National Council on Radiation Protection and Measurements, http://NCRPonline.org . Managing Space Radiation Risk in the New Era of Space Exploration (2008). Committee on the Evaluation of Radiation Shielding for Space Exploration, National Research Council. National Academies Press, pages 111-118. -A- AAPM: American Association of Physicists in Medicine. absolute risk: Expression of excess risk due to exposure as the arithmetic difference between the risk among those exposed and that obtaining in the absence of exposure. absorbed dose (D): Average amount of energy imparted by ionizing particles to a unit mass of irradiated material in a volume sufficiently small to disregard variations in the radiation field but sufficiently large to average over statistical fluctuations in energy deposition, and where energy imparted is the difference between energy entering the volume and energy leaving the volume. The same dose has different consequences depending on the type of radiation delivered. Unit: gray (Gy), equivalent to 1 J/kg. ACE: Advanced Composition Explorer Mission, launched in 1997 and orbiting the L1 libration point to sample energetic particles arriving from the Sun and interstellar and galactic sources. It also provides continuous coverage of solar wind parameters and solar energetic particle intensities (space weather). When reporting space weather, it can provide an advance warning (about one hour) of geomagnetic storms that can overload power grids, disrupt communications on Earth, and present a hazard to astronauts.
    [Show full text]
  • Radiation Risk in Perspective
    PS010-1 RADIATION RISK IN PERSPECTIVE POSITION STATEMENT OF THE HEALTH HEALTH PHYSICS SOCIETY* PHYSICS SOCIETY Adopted: January 1996 Revised: August 2004 Contact: Richard J. Burk, Jr. Executive Secretary Health Physics Society Telephone: 703-790-1745 Fax: 703-790-2672 Email: [email protected] http://www.hps.org In accordance with current knowledge of radiation health risks, the Health Physics Society recommends against quantitative estimation of health risks below an individual dose of 5 rem1 in one year or a lifetime dose of 10 rem above that received from natural sources. Doses from natural background radiation in the United States average about 0.3 rem per year. A dose of 5 rem will be accumulated in the first 17 years of life and about 25 rem in a lifetime of 80 years. Estimation of health risk associated with radiation doses that are of similar magnitude as those received from natural sources should be strictly qualitative and encompass a range of hypothetical health outcomes, including the possibility of no adverse health effects at such low levels. There is substantial and convincing scientific evidence for health risks following high-dose exposures. However, below 5–10 rem (which includes occupational and environmental exposures), risks of health effects are either too small to be observed or are nonexistent. In part because of the insurmountable intrinsic and methodological difficulties in determining if the health effects that are demonstrated at high radiation doses are also present at low doses, current radiation protection standards and practices are based on the premise that any radiation dose, no matter how small, may result in detrimental health effects, such as cancer and hereditary genetic damage.
    [Show full text]
  • Experimental Γ Ray Spectroscopy and Investigations of Environmental Radioactivity
    Experimental γ Ray Spectroscopy and Investigations of Environmental Radioactivity BY RANDOLPH S. PETERSON 216 α Po 84 10.64h. 212 Pb 1- 415 82 0- 239 β- 01- 0 60.6m 212 1+ 1630 Bi 2+ 1513 83 α β- 2+ 787 304ns 0+ 0 212 α Po 84 Experimental γ Ray Spectroscopy and Investigations of Environmental Radioactivity Randolph S. Peterson Physics Department The University of the South Sewanee, Tennessee Published by Spectrum Techniques All Rights Reserved Copyright 1996 TABLE OF CONTENTS Page Introduction ....................................................................................................................4 Basic Gamma Spectroscopy 1. Energy Calibration ................................................................................................... 7 2. Gamma Spectra from Common Commercial Sources ........................................ 10 3. Detector Energy Resolution .................................................................................. 12 Interaction of Radiation with Matter 4. Compton Scattering............................................................................................... 14 5. Pair Production and Annihilation ........................................................................ 17 6. Absorption of Gammas by Materials ..................................................................... 19 7. X Rays ..................................................................................................................... 21 Radioactive Decay 8. Multichannel Scaling and Half-life .....................................................................
    [Show full text]
  • Cherenkov Radiation
    TheThe CherenkovCherenkov effecteffect A charged particle traveling in a dielectric medium with n>1 radiates Cherenkov radiation B Wave front if its velocity is larger than the C phase velocity of light v>c/n or > 1/n (threshold) A β Charged particle The emission is due to an asymmetric polarization of the medium in front and at the rear of the particle, giving rise to a varying electric dipole momentum. dN Some of the particle energy is convertedγ = 491into light. A coherent wave front is dx generated moving at velocity v at an angle Θc If the media is transparent the Cherenkov light can be detected. If the particle is ultra-relativistic β~1 Θc = const and has max value c t AB n 1 cosθc = = = In water Θc = 43˚, in ice 41AC˚ βct βn 37 TheThe CherenkovCherenkov effecteffect The intensity of the Cherenkov radiation (number of photons per unit length of particle path and per unit of wave length) 2 2 2 2 2 Number of photons/L and radiation d N 4π z e 1 2πz 2 = 2 1 − 2 2 = 2 α sin ΘC Wavelength depends on charge dxdλ hcλ n β λ and velocity of particle 2πe2 α = Since the intensity is proportional to hc 1/λ2 short wavelengths dominate dN Using light detectors (photomultipliers)γ = sensitive491 in 400-700 nm for an ideally 100% efficient detector in the visibledx € 2 dNγ λ2 d Nγ 2 2 λ2 dλ 2 2 11 1 22 2 d 2 z sin 2 z sin 490393 zz sinsinΘc photons / cm = ∫ λ = π α ΘC ∫ 2 = π α ΘC 2 −− 2 = α ΘC λ1 λ1 dx dxdλ λ λλ1 λ2 d 2 N d 2 N dλ λ2 d 2 N = = dxdE dxdλ dE 2πhc dxdλ Energy loss is about 104 less hc 2πhc than 2 MeV/cm in water from €
    [Show full text]
  • Nuclide Safety Data Sheet Iron-59
    59 Nuclide Safety Data Sheet 59 Iron-59 Fe www.nchps.org Fe I. PHYSICAL DATA 1 Gammas/X-rays: 192 keV (3% abundance); 1099 keV (56%), 1292 keV (44%) Radiation: 1 Betas: 131 keV (1%), 273 keV (46%), 466 keV (53%) 2 Gamma Constant: 0.703 mR/hr per mCi @ 1.0 meter [1.789E-4 mSv/hr per MBq @ 1.0 meter] 1 Half-Life [T½] : Physical T½: 44.5 day 3 Biological T½: ~1.7 days (ingestion) ; much longer for small fraction Effective T½: ~1.7 days (varies; also small fraction retained much longer) Specific Activity: 4.97E5 Ci/g [1.84E15 Bq/g] max.1 II. RADIOLOGICAL DATA Radiotoxicity4: 1.8E-9 Sv/Bq (67 mrem/uCi) of 59Fe ingested [CEDE] 4.0E-9 Sv/Bq (150 mrem/uCi) of 59Fe inhaled [CEDE] Critical Organ: Spleen1, [blood] Intake Routes: Ingestion, inhalation, puncture, wound, skin contamination (absorption); Radiological Hazard: Internal Exposure; Contamination III. SHIELDING Half Value Layer [HVL] Tenth Value Layer [TVL] Lead [Pb] 15 mm (0.59 inches) 45 mm (1.8 inches) Steel 35 mm (1.4 inches) 91 mm (3.6 inches) The accessible dose rate should be background but must be < 2 mR/hr IV. DOSIMETRY MONITORING - Always wear radiation dosimetry monitoring badges [body & ring] whenever handling 59Fe V. DETECTION & MEASUREMENT Portable Survey Meters: Geiger-Mueller [e.g. PGM] to assess shielding effectiveness & contamination Wipe Test: Liquid Scintillation Counter, Gamma Counter VI. SPECIAL PRECAUTIONS • Avoid skin contamination [absorption], ingestion, inhalation, & injection [all routes of intake] • Use shielding (lead) to minimize exposure while handling of 59Fe • Use tools (e.g.
    [Show full text]
  • 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.
    [Show full text]
  • 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.
    [Show full text]
  • 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
    [Show full text]
  • Occupational Radiation Protection Record-Keeping and Reporting Guide
    DOE G 441.1-11 (formerly G-10 CFR 835/H1) 05-20-99 OCCUPATIONAL RADIATION PROTECTION RECORD-KEEPING AND REPORTING GUIDE for use with Title 10, Code of Federal Regulations, Part 835, Occupational Radiation Protection Assistant Secretary for Environment, Safety and Health (THIS PAGE INTENTIONALLY LEFT BLANK) DOE G 441.1-11 i 05-20-99 CONTENTS CONTENTS PAGE 1. PURPOSE AND APPLICABILITY ........................................................ 1 2. DEFINITIONS ........................................................................ 2 3. DISCUSSION ........................................................................ 3 4. IMPLEMENTATION GUIDANCE ......................................................... 4 4.1 RECORDS TO BE GENERATED AND MAINTAINED ................................ 4 4.1.1 Individual Monitoring and Dose Records ........................................ 4 4.1.2 Monitoring and Workplace Records ............................................ 8 4.1.3 Administrative Records .................................................... 11 4.2 REPORTS ................................................................... 15 4.2.1 Reports to Individuals ...................................................... 16 4.2.2 Reports of Planned Special Exposures ......................................... 17 4.3 PRIVACY ACT CONSIDERATIONS .............................................. 17 4.3.1 Informing Individuals ...................................................... 17 4.3.2 Identifying Individuals ..................................................... 17
    [Show full text]
  • 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 .
    [Show full text]
  • 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).
    [Show full text]