Management of Persons Contaminated With Radionuclides: Handbook

NCRP REPORT No. 161

MANAGEMENT OF PERSONS CONTAMINATED WITH RADIONUCLIDES: HANDBOOK

NCRP REPORT No. 161 I

Management of Persons Contaminated with Radionuclides: Handbook

Recommendations of the NATIONAL COUNCIL ON RADIATION PROTECTION AND MEASUREMENTS

December 20, 2008

National Council on Radiation Protection and Measurements 7910 Woodmont Avenue, Suite 400 / Bethesda, MD 20814-3095

LEGAL NOTICE This Report was prepared by the National Council on Radiation Protection and Measurements (NCRP). The Council strives to provide accurate, complete and useful information in its documents. However, neither NCRP, the members of NCRP, other persons contributing to or assisting in the preparation of this Report, nor any person acting on the behalf of any of these parties: (a) makes any warranty or representation, express or implied, with respect to the accuracy, completeness or usefulness of the information contained in this Report, or that the use of any information, method or process disclosed in this Report may not infringe on privately owned rights; or (b) assumes any liability with respect to the use of, or for damages resulting from the use of any information, method or process disclosed in this Report, under the Civil Rights Act of 1964, Section 701 et seq. as amended 42 U.S.C. Section 2000e et seq. (Title VII) or any other statutory or common law theory governing liability.

Disclaimer Any mention of commercial products within NCRP publications is for information only; it does not imply recommendation or endorsement by NCRP.

Library of Congress Cataloging-in-Publication Data National Council on Radiation Protection and Measurements. Management of persons contaminated with radionuclides : recommendations of the National Council on Radiation Protection and Measurements, December 20, 2008. p. ; cm. -- (NCRP report ; no. 161) Includes bibliographical references and index. ISBN-13: 978-0-929600-99-4 ISBN-10: 0-929600-99-1 1. Radiation injuries. 2. Radioisotopes in the body. 3. Environmental toxicology. 4. First aid in illness and injury. I. Title. II. Series: NCRP report ; no. 161. [DNLM: 1. Radiation Injuries--therapy. 2. Environmental Exposure. 3. First Aid. 4. Radioisotopes--adverse effects. WN 610 M266 2009] RC93.N37 2009 362.196'9897--dc22 2009045936

Copyright © National Council on Radiation Protection and Measurements 2009 All rights reserved. This publication is protected by copyright. No part of this publication may be reproduced in any form or by any means, including photocopying, or utilized by any information storage and retrieval system without written permission from the copyright owner, except for brief quotation in critical articles or reviews.

Preface The National Council on Radiation Protection and Measurements (NCRP) published Report No. 65 on Management of Persons Accidentally Contaminated with Radionuclides in 1980. This Report has served as a major resource for responders to accidents and incidents involving human contamination by radionuclides. During the last three decades a greater understanding has been achieved on the possible health effects in, and strategies for the immediate and late management of, contaminated individuals. In recent years, the range of situations in which contamination can occur has increased with the growing concern worldwide regarding possible incidents of nuclear and radiological terrorism. At the time of publication of NCRP Report No. 65, the main concern was the possible contamination of individuals working at, or living near, a nuclear-reactor facility and those working at, or living near, the government’s nuclear-weapons sites. This concern has now expanded into the broader public domain and involves a greater range of radionuclides than those of greatest concern in an incident involving nuclear-reactor operations, a reactor accident, or an accidental release of radionuclides at a weapons site. This Report therefore has been significantly extended beyond the set of radionuclides that were considered in Report No. 65, and contains recommendations on the management of persons contaminated by many radionuclides of concern in potential acts of nuclear or radiological terrorism. It also provides information based on advances since the 1970s in methods for decontamination and the decorporation of radionuclides in accidentally or deliberately contaminated persons. For example, the Report includes updated data and biokinetic and dosimetric models of organ doses, total-body and organ retention values, and excretion rates of radionuclides. Publications of the International Commission on Radiological Protection over the past three decades have provided valuable information that is utilized in this Report. The Report contains five major sections: 1. Part A is an update of the “yellow” section of NCRP Report No. 65 and contains quick reference information needed by an emergency responder to an act of radionuclide contamination; iii

iv / PREFACE 2. Part B contains a set of recommendations on onsite and prehospital actions that should be taken by responders; 3. Part C contains an extensive discussion of actions that should be taken in the treatment of contaminated patients at a medical facility; 4. Part D contains recommendations on post-treatment followup and guidance on contamination control in handling decedents; and 5. Part E provides an in-depth discussion of the scientific and technical bases for the recommended management procedures for individuals contaminated with radionuclides, including detailed discussions of internal dosimetry models for major radionuclides of 24 elements of particular concern. Parts A, B, C and D are presented separately as a handbook for the convenience of users who might want to have the information readily available at an incident site. Part E is presented in the second volume. Volume 2 (Sections 16 through 22 and Appendices A to J) of Report No. 161 contains extensive information on the Scientific and Technical Bases for the guidance provided in Volume 1. Included are a detailed presentation on the radiobiology of internally-deposited radionuclides, a discussion of sources of potential contamination in both planned (e.g., medical or industrial) and unplanned (e.g., nuclear accidents or acts of terrorism) settings, roles and responsibilities of responders to incidents involving radionuclide contamination, extensive dosimetry and case studies for radionuclides of 24 important chemical elements, and guidance on dose assessment methodologies. Both volumes of Report No. 161 were prepared by Scientific Committee 4-1. Serving on the Committee were: William J Bair, Chairman Pacific Northwest National Laboratory (retired) Richland, Washington Members Wesley E. Bolch University of Florida Gainesville, Florida

William E. Dickerson Armed Forces Radiobiology Research Institute Bethesda, Maryland

PREFACE

Keith F. Eckerman Oak Ridge National Laboratory Oak Ridge, Tennessee

Ronald E. Goans MJW Corporation Clinton, Tennessee

P. Andrew Karam New York City Department of Health and Mental Hygiene New York, New York

Richard W. Leggett Oak Ridge National Laboratory Oak Ridge, Tennessee

Joyce L. Lipsztein State University of Rio de Janeiro Rio de Janeiro, Brazil

Michael G. Stabin Vanderbilt University Nashville, Tennessee

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Albert L. Wiley, Jr. Radiation Emergency Assistance Center/Training Site Oak Ridge, Tennessee

Consultants Eugene H. Carbaugh Bryce D. Breitenstein, Jr. Brookhaven National Laboratory Pacific Northwest National Laboratory (retired) Richland, Washington Long Beach, California

NCRP Secretariat Bruce B. Boecker, Staff Consultant Cindy L. O’Brien, Managing Editor David A. Schauer, Executive Director

NCRP acknowledges and thanks the U.S. Navy, the U.S. Nuclear Regulatory Commission, and the Centers for Disease Control and Prevention for providing funds to support the preparation of this Report. The Council also expresses appreciation to the members of Committee 4-1, who invested great effort and personal time in the preparation of the Report and thanks William J Bair, III, National Security Technologies, LLC, Nevada Test Site, Mercury, Nevada, for his contributions to the operational health-physics aspects of the Report. Thomas S. Tenforde President

Contents Volume I Preface . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii 1. Executive Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1 2. Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .11 2.1 Purpose of this Report . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12 2.2 Target Audiences of this Report . . . . . . . . . . . . . . . . . . . . .15 2.3 Organization of this Report . . . . . . . . . . . . . . . . . . . . . . . .15 2.3.1 Management of Persons Contaminated with Radionuclides: Handbook . . . . . . . . . . . . . . . . . . .16 2.3.2 Part A: Quick Reference Information . . . . . . . . . .19 2.3.3 Part B: Onsite and Prehospital Actions . . . . . . . .19 2.3.4 Part C: Patient Management at Hospital . . . . . . .20 2.3.5 Part D: Patient Management Post-Hospital . . . .20 2.3.6 Management of Persons Contaminated with Radionuclides: Scientific and Technical Bases . . .21 Part A: Quick Reference Information 3. Compendium of Radiation Facts and Guidance . . . . . . . .25 3.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .25 3.1.1 Organizations Offering Radiological Incident Assistance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .25 3.1.2 Terminology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .26 3.2 Basic Radiological Facts . . . . . . . . . . . . . . . . . . . . . . . . . . .30 3.2.1 Radiation Types and Recommended Personnel Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .30 3.2.2 Identifying Radiation Types Using a Pancake or Other Thin End-Window Geiger-Mueller Probe Survey Meter . . . . . . . . . . . . . . . . . . . . . . . . . . . . .31 3.2.3 Radiation Energy and Radioactive Decay Facts .32 3.3 Incident Response (Section 18) . . . . . . . . . . . . . . . . . . . . . .33 3.3.1 Incidents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .33 3.3.1.1 Small Scale . . . . . . . . . . . . . . . . . . . . . . .33 3.3.1.2 Large Scale . . . . . . . . . . . . . . . . . . . . . . .34 3.3.2 Roles and Responsibilities (Section 18) . . . . . . . .34 3.4 Guidance for Professionals at Incident Site . . . . . . . . . . . .35

vii

viii / CONTENTS 3.4.1

Radiation Readings and Their Significance (dose-rate meters) . . . . . . . . . . . . . . . . . . . . . . . . . 36 3.4.2 Surface Radiation Readings and Their Significance (contamination survey meters) . . . . . . . . . . . . . . 37 3.5 Management of Potentially-Injured and Contaminated Persons . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 3.5.1 Priorities for Aiding Contaminated Individuals . 38 3.5.2 Stages in Management of Exposed Persons . . . . 38 3.6 Radiation Exposures from External Sources . . . . . . . . . . 42 3.6.1 Health Effects from External Radiation . . . . . . . 42 3.6.2 Neutron-Radiation Dose from Criticality Accident (based on 24Na activation in body) . . . . . . . . . . . . 42 3.6.3 Exposures from Sealed Radioactive Sources . . . 45 3.7 Air Kerma and Skin Doses for Point Sources . . . . . . . . . . 46 3.7.1 Intervention Levels for Skin Contamination . . . 46 3.7.2 Guidance for Decontaminating Skin . . . . . . . . . . 49 3.8 Radiation Exposures from Internal Depositions of Radionuclides . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49 3.8.1 Health Effects from Internal Radionuclide Contamination . . . . . . . . . . . . . . . . . . . . . . . . . . . 49 3.8.1.1 Deterministic Effects (harmful tissue reactions) . . . . . . . . . . . . . . . . . . 49 3.8.1.2 Stochastic Effects . . . . . . . . . . . . . . . . . 52 3.8.1.2.1 Cancer . . . . . . . . . . . . . . . . . . 52 3.8.1.2.2 Hereditary effects (Section 16.7.2.2). . . . . . . . . . 52 3.8.1.3 Developmental Effects (Section 16.7.3) . . . . . . . . . . . . . . . . . . . 52 3.8.2 Inhalation Intakes . . . . . . . . . . . . . . . . . . . . . . . . 54 3.8.2.1 Air Samples . . . . . . . . . . . . . . . . . . . . . . 54 3.8.2.2 Nasal Swabs . . . . . . . . . . . . . . . . . . . . . 54 3.8.2.3 Doses Received from Inhaled Radionuclides . . . . . . . . . . . . . . . . . . . . 56 3.8.3 Intakes Through Skin and Ingestion . . . . . . . . . 56 3.9 Medical Management of Internal Radionuclide Depositions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56 3.9.1 Clinical Decision Guides . . . . . . . . . . . . . . . . . . . 56 3.9.2 Decorporation Therapy . . . . . . . . . . . . . . . . . . . . 63 3.10 Radiation Dose Limitation . . . . . . . . . . . . . . . . . . . . . . . . 63 4. Radiation-Safety Guidance for First Responders . . . . . . 4.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.2 General Instructions for First Responders . . . . . . . . . . . . 4.3 Guidance for First Responders . . . . . . . . . . . . . . . . . . . . . 4.3.1 First on the Scene . . . . . . . . . . . . . . . . . . . . . . . . .

72 72 73 74 75

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4.3.2 4.3.3 4.3.4

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Immediate Goals for Protection of Exposed Individuals . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .75 Control Areas . . . . . . . . . . . . . . . . . . . . . . . . . . . .76 Protection of First Responders . . . . . . . . . . . . . . .78

5. Performing Surveys and Controlling Personnel and Area Contamination . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .80 5.1 Contamination Surveys . . . . . . . . . . . . . . . . . . . . . . . . . . .80 5.1.1 How to Perform Surveys of Individuals, Clothing, Samples and Surfaces . . . . . . . . . . . . . . . . . . . . . .81 5.1.2 How to Perform a Beta/Gamma-Radiation Area Survey . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .82 5.1.3 How to Perform an Alpha-Radiation Survey of Contaminated Areas, Individuals and Samples .82 5.2 Personal Protection Equipment . . . . . . . . . . . . . . . . . . . . .83 5.2.1 Examples of Personal Protection Equipment . . . .83 5.2.2 Personal Protection Equipment Inspection . . . . .83 5.2.3 Dressing in Personal Protection Equipment . . . .86 5.2.4 Removing Personal Protection Equipment . . . . .86 5.2.5 Actions to be Taken after Personal Protection Equipment is Removed . . . . . . . . . . . . . . . . . . . . .88 5.3 Contamination Control . . . . . . . . . . . . . . . . . . . . . . . . . . . .88 5.3.1 Contamination Control Practices . . . . . . . . . . . . .88 5.3.2 Contamination Control of Exposed People . . . . . .88 5.3.3 Contamination Control Among Medical and Emergency- Response Personnel . . . . . . . . . . . . .89 5.3.4 Radiologically-Controlled Areas (further defined in Sections 4.3.3 and 18) . . . . . . . . . . . . . . . . . . . .89 5.3.4.1 General Guidelines for Operation of a Controlled Contamination Area . . . . . .90 5.3.4.2 Leaving a Controlled Area . . . . . . . . . . .91 5.3.4.3 Transportation of Injured and Contaminated Individuals. . . . . . . . . . . 91 5.3.5 Decontamination of Equipment . . . . . . . . . . . . . .92 Part B: Onsite and Prehospital Actions 6. Stage 1: Medical Assessment (onsite triage area) . . . . . . .93 6.1 Initial Actions of Medical and Radiation Safety Personnel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .94 6.2 Potential Life-Threatening Problems . . . . . . . . . . . . . . . . .94 6.3 Identification of Individuals Exposed to Radiation and/or Radionuclides . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .96 6.4 Assessment of External Irradiation . . . . . . . . . . . . . . . . . .96 6.5 Contamination Screening of Individuals . . . . . . . . . . . . .101 6.5.1 External Contamination . . . . . . . . . . . . . . . . . . .101

x / CONTENTS 6.5.2

Internal Contamination . . . . . . . . . . . . . . . . . . . 6.5.2.1 Inhalation Intakes . . . . . . . . . . . . . . . 6.5.2.2 Intakes Through Skin and Ingestion . 6.5.2.3 Collection of Excreta . . . . . . . . . . . . . . Onsite Treatment for Internal Contamination . . . . . . . Priorities in Processing Exposed Persons . . . . . . . . . . . . Documenting a Radionuclide Contamination Incident .

101 102 103 103 104 104 104

7. Stage 2: External Contamination Assessment (onsite triage area) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.1 External Contamination Assessment Procedures . . . . . 7.2 Dose Assessment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.3 Screening Guidance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.4 Treatment Guidance . . . . . . . . . . . . . . . . . . . . . . . . . . . .

107 107 109 111 111

6.6 6.7 6.8

8. Stage 3: External Decontamination (onsite decontamination area) . . . . . . . . . . . . . . . . . . . . . . 8.1 Decontamination of Persons . . . . . . . . . . . . . . . . . . . . . . 8.1.1 Decontamination Objectives . . . . . . . . . . . . . . . 8.1.2 Decontamination Procedures . . . . . . . . . . . . . . . 8.2 Guidance for Those Performing Decontamination Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.3 Decontamination Facilities . . . . . . . . . . . . . . . . . . . . . . . 8.4 Saving Contaminated Materials . . . . . . . . . . . . . . . . . . . 8.5 Management of Individuals After Contamination Assessment and Decontamination of Skin and Wounds . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.6 Documentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Part C: Patient Management at Hospital 9. Stage 4: Patient Evaluation and Emergency Care (hospital) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9.1 General Issues in Initial Patient Evaluation . . . . . . . . . 9.1.1 Medical Evaluation of Persons with Internal Contamination . . . . . . . . . . . . . . . . . . . . . . . . . . 9.1.2 Psychological and Behavioral Consequence Management After Radiation Incidents . . . . . . 9.2 General Instructions for Emergency Department Medical Staff . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9.3 Emergency Medical Management . . . . . . . . . . . . . . . . . . 9.3.1 Caring for Contaminated Individuals with Life-Threatening Injuries . . . . . . . . . . . . . . . . . . 9.3.2 Caring for Lightly Injured and Uninjured Contaminated Exposed Persons . . . . . . . . . . . .

113 113 114 115 117 118 120

120 121

123 124 124 125 127 128 128 128

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9.3.3

9.4

9.5

Caring for Persons Suffering from Radiation Injury . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .129 9.3.4 Caring for Persons Suffering from Radiation Burns . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .131 Initial Treatment Decisions . . . . . . . . . . . . . . . . . . . . . . .132 9.4.1 Radionuclides in a Wound . . . . . . . . . . . . . . . . . .132 9.4.2 Radionuclide Inhalation . . . . . . . . . . . . . . . . . . .133 9.4.3 Radionuclide Ingestion . . . . . . . . . . . . . . . . . . . .134 9.4.3.1 Gastric Lavage . . . . . . . . . . . . . . . . . . .134 9.4.3.2 Emetics . . . . . . . . . . . . . . . . . . . . . . . . .135 9.4.3.3 Purgatives. . . . . . . . . . . . . . . . . . . . . . .135 9.4.4 Clinical Decision Guides . . . . . . . . . . . . . . . . . .136 9.4.5 Specific Drug Decorporation Therapy . . . . . . . .136 9.4.6 Algorithm for Medical Management of Internal Depositions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .136 Medical Information Checklist . . . . . . . . . . . . . . . . . . . . .136

10. Stage 5: Internal Contamination Assessment (hospital) 138 10.1 Preliminary Assessment Activities . . . . . . . . . . . . . . . . .138 10.2 Information About the Contaminating Incident . . . . . . .140 10.2.1 Location of the Individual and Time of Exposure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .140 10.2.2 Establishing the Route of Exposure . . . . . . . . . .141 10.2.3 Radionuclide Identification and Physical and Chemical Form . . . . . . . . . . . . . . . . . . . . . . . . . . .142 10.3 Bioassay . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .143 10.3.1 Indirect (in vitro) Bioassay Monitoring . . . . . . .144 10.3.1.1 Nasal Swabs . . . . . . . . . . . . . . . . . . . . .144 10.3.1.2 Urine Bioassay . . . . . . . . . . . . . . . . . . .146 10.3.1.3 Fecal Samples . . . . . . . . . . . . . . . . . . . .150 10.3.1.4 Blood Samples . . . . . . . . . . . . . . . . . . .150 10.3.1.5 Tissue Specimens . . . . . . . . . . . . . . . . .151 10.3.2 Direct (in vivo) Monitoring . . . . . . . . . . . . . . . . .151 10.3.2.1 Whole-Body Counting . . . . . . . . . . . . .152 10.3.2.2 Chest (lung) Counting . . . . . . . . . . . . .152 10.3.2.3 Counting of Particular Organs or Tissues . . . . . . . . . . . . . . . . . . . . . . . . .152 10.4 Intake and Dose Assessment . . . . . . . . . . . . . . . . . . . . . .153 11. Stage 6: Clinical Decision Guidance (hospital) . . . . . . . .158 11.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .158 11.2 Clinical Decision Guides . . . . . . . . . . . . . . . . . . . . . . . . . .158 11.3 Clinical Decision Guide Instrument Considerations . . . .169 11.4 Extrapolation of Data for Spot Urine Sample to 24 h Values . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .169 11.5 Worked Examples with Bioassay Data . . . . . . . . . . . . . .171

xii / CONTENTS 12. Stage 7: Medical Management (hospital) . . . . . . . . . . . . . 12.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.2 Decorporation Therapy for Internally-Deposited Radionuclides . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.3 Decorporation Therapy by Drug . . . . . . . . . . . . . . . . . . . 12.3.1 Deferoxamine Treatment . . . . . . . . . . . . . . . . . . 12.3.2 Dimercaprol Treatment . . . . . . . . . . . . . . . . . . . 12.3.3 DTPA Treatment . . . . . . . . . . . . . . . . . . . . . . . . 12.3.4 Ethylenediaminetetraacetic Acid Treatment . . 12.3.5 Penicillamine Treatment . . . . . . . . . . . . . . . . . . 12.3.6 Prussian Blue Insoluble Treatment . . . . . . . . . 12.3.7 Succimer Treatment . . . . . . . . . . . . . . . . . . . . . . 12.4 Medical Treatments Arranged by Radionuclide . . . . . . . 12.4.1 Medical Treatment for Barium and Calcium Radionuclides . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.4.2 Medical Treatment for Cobalt Radionuclides . . 12.4.3 Medical Treatment for Iodine Radionuclides . . 12.4.4 Medical Treatment for Radioactive Phosphorus 12.4.5 Medical Treatment for Radium and Strontium Radionuclides . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.4.6 Medical Treatment for Tritium (radioactive hydrogen) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.4.7 Medical Treatment for Uranium Isotopes . . . . . 12.4.8 Medical Treatment for the Actinide Nuclides (Section 12.3.3) . . . . . . . . . . . . . . . . . . . . . . . . . . 12.5 Lung Lavage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

176 177 179 187 187 189 191 198 199 201 209 210 210 212 212 221 221 228 228 233 233

Part D: Patient Management Post-Hospital 13. Stage 8: Follow-Up Medical Care . . . . . . . . . . . . . . . . . . . . 13.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13.2 Late-Occurring Health Effects . . . . . . . . . . . . . . . . . . . . 13.3 Preventive Medicine Approaches . . . . . . . . . . . . . . . . . . 13.4 Psychosocial Issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

237 237 238 239 240

14. Stage 9: Contaminated Decedents (hospital and mortuary) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2 Guidelines for the Medical Examiner . . . . . . . . . . . . . . . 14.2.1 Field Activities . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2.2 Autopsy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.3 Guidelines for Mortuary Personnel . . . . . . . . . . . . . . . . . 14.4 Final Disposition of the Decedent . . . . . . . . . . . . . . . . . . 14.5 Religious and Cultural Considerations . . . . . . . . . . . . . .

244 244 245 245 247 251 252 253

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15. Contamination Control in Medical Facilities . . . . . . . . . .255 15.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .255 15.2 Standard Precautions . . . . . . . . . . . . . . . . . . . . . . . . . . . .255 15.3 Contamination Control Actions in the Emergency Department for Highly-Contaminated Patients . . . . . . .256 15.4 Working with Contaminated Patients . . . . . . . . . . . . . . .257 15.5 Hospital Emergency Department Contamination Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .257 15.6 Contaminated Patients’ Rooms . . . . . . . . . . . . . . . . . . . .259 15.7 Patient Decontamination (Section 8) . . . . . . . . . . . . . . . .259 15.8 Responsibilities of Radiation-Safety Personnel . . . . . . . .260 15.9 Hospital Decontamination Procedures for Protection of Personnel and Facilities . . . . . . . . . . . . . . . . . . . . . . . . . .260 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .263 Index . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .277

Volume II Preface . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii 16. Overview of Radiobiology Concepts Pertinent to Radionuclides . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .287 16.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .287 16.2 Radionuclides and Radiation . . . . . . . . . . . . . . . . . . . . . .288 16.3 General Characteristics of Radionuclide and Radiation Exposures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .293 16.4 Radionuclides as Internal Health Risks . . . . . . . . . . . . .296 16.5 Routes of Entry into the Body . . . . . . . . . . . . . . . . . . . . .297 16.5.1 Skin and Wound Contamination. . . . . . . . . . . .298 16.5.2 Inhalation of Radionuclides. . . . . . . . . . . . . . . .300 16.5.3 Ingestion of Radionuclides. . . . . . . . . . . . . . . . .308 16.6 Internal Dosimetry Models . . . . . . . . . . . . . . . . . . . . . . .310 16.7 Quantities Used in Radiation Protection . . . . . . . . . . . .312 16.7.1 Units of Activity in Current Use . . . . . . . . . . .312 16.7.2 Dosimetric Quantities and Units in Current Use . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .313 16.8 Health Effects from Radiation Exposures . . . . . . . . . . .315 16.8.1 Deterministic Effects . . . . . . . . . . . . . . . . . . . .319 16.8.1.1 Thresholds and Unrecoverable Levels for Deterministic Effects. . .319 16.8.1.2 Dependence of Threshold and D50 Values on Dose Rate . . . . . . . . . . . 324

16.8.1.3 16.8.1.4 16.8.1.5

16.9

Acute Radiation Syndrome . . . . . . Relative Biological Effectiveness . Deterministic Effects of Concern Following Intakes of Radionuclides. . . . . . . . . . . . . . . . . 16.8.1.6 Beta Burns . . . . . . . . . . . . . . . . . . . 16.8.1.7 Hematopoietic (bone-marrow) Failure . . . . . . . . . . . . . . . . . . . . . . 16.8.1.8 Impaired Pulmonary Function . . . 16.8.1.9 Gastrointestinal System Failure. . 16.8.1.10 Thyroid . . . . . . . . . . . . . . . . . . . . . . 16.8.1.11 Summary of Deterministic Health Effects. . . . . . . . . . . . . . . . . . . . . . . 16.8.2 Stochastic Effects . . . . . . . . . . . . . . . . . . . . . . . 16.8.2.1 Cancer . . . . . . . . . . . . . . . . . . . . . . 16.8.2.2 Hereditary Effects . . . . . . . . . . . . . 16.8.3 Developmental Effects . . . . . . . . . . . . . . . . . . . 16.8.4 Hot Particles . . . . . . . . . . . . . . . . . . . . . . . . . . Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16.9.1 Use of Dose Quantities and Units . . . . . . . . . . 16.9.2 Deterministic and Stochastic Health Effects .

17. Settings in Which Individuals May be Contaminated with Radionuclides . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17.2 Types of Contamination Incidents . . . . . . . . . . . . . . . . . 17.2.1 Small Scale . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17.2.2 Large Scale . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17.2.3 Accidental . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17.2.4 Deliberate . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17.3 Settings in Which Contamination Incidents May Occur . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17.3.1 Malicious Use: Stolen or Improvised Nuclear Device . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17.3.1.1 Examples of Contamination Incidents. . . . . . . . . . . . . . . . . . . . . 17.3.1.2 Likely Radionuclides Involved . . . 17.3.1.3 Radiological Considerations . . . . . 17.3.1.4 Important Nonradiological Considerations . . . . . . . . . . . . . . . . 17.3.1.5 Examples and Descriptions of Sources and Devices . . . . . . . . . . . 17.3.2 Malicious Use: Radiological Dispersal Device 17.3.2.1 Examples of Contamination Incidents. . . . . . . . . . . . . . . . . . . . .

324 325

326 327 329 330 338 339 339 341 343 356 358 359 360 360 361

363 363 363 363 364 365 366 367 391 392 392 392 392 393 393 394

CONTENTS

17.3.2.2 17.3.2.3

17.3.3

17.3.4

17.3.5

17.3.6

17.3.7

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Radiological Considerations . . . . . .394 Important Nonradiological Considerations . . . . . . . . . . . . . . . .396 17.3.2.4 Examples and Descriptions of Sources and Devices . . . . . . . . . . . .396 Malicious Use: Contamination of Food or Water Supply . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .396 17.3.3.1 Examples of Contamination Incidents . . . . . . . . . . . . . . . . . . . . .397 17.3.3.2 Likely Radionuclides Involved . . . .397 17.3.3.3 Radiological Considerations . . . . . .397 17.3.3.4 Important Nonradiological Considerations . . . . . . . . . . . . . . . .398 17.3.3.5 Examples and Descriptions of Sources and Devices . . . . . . . . . . . .398 Malicious Use: Deliberate Contamination of Another Individual . . . . . . . . . . . . . . . . . . . . . .398 17.3.4.1 Examples of Contamination Incidents . . . . . . . . . . . . . . . . . . . . .399 17.3.4.2 Likely Radionuclides Involved . . . .399 17.3.4.3 Radiological Considerations . . . . . .399 17.3.4.4 Important Nonradiological Considerations . . . . . . . . . . . . . . . .400 17.3.4.5 Examples and Descriptions of Sources and Devices . . . . . . . . . . . .400 Nuclear Reactors . . . . . . . . . . . . . . . . . . . . . . .400 17.3.5.1 Examples of Contamination Incidents . . . . . . . . . . . . . . . . . . . . .401 17.3.5.2 Radiological Considerations . . . . . .401 17.3.5.3 Important Nonradiological Considerations . . . . . . . . . . . . . . . .404 17.3.5.4 Examples and Descriptions of Sources and Devices . . . . . . . . . . . .405 Industrial: Source Manufacture . . . . . . . . . . . .405 17.3.6.1 Examples of Contamination Incidents . . . . . . . . . . . . . . . . . . . . .406 17.3.6.2 Likely Radionuclides Involved . . . .406 17.3.6.3 Radiological Considerations . . . . . .406 17.3.6.4 Important Nonradiological Considerations . . . . . . . . . . . . . . . .407 Industrial: Source Use . . . . . . . . . . . . . . . . . . .407 17.3.7.1 Examples of Contamination Incidents . . . . . . . . . . . . . . . . . . . . .408 17.3.7.2 Radiological Considerations . . . . . .408

xvi / CONTENTS 17.3.7.3

Important Nonradiological Considerations . . . . . . . . . . . . . . . . 410 17.3.7.4 Examples and Descriptions of Sources and Devices . . . . . . . . . . . 410 17.3.8 Radioactive-Material Transportation . . . . . . . 410 17.3.8.1 Examples of Contamination Incidents. . . . . . . . . . . . . . . . . . . . . 411 17.3.8.2 Likely Radionuclides Involved . . . 411 17.3.8.3 Radiological Considerations . . . . . 411 17.3.8.4 Important Nonradiological Considerations . . . . . . . . . . . . . . . . 411 17.3.8.5 Examples and Descriptions of Sources and Devices . . . . . . . . . . . 412 17.3.9 Medical . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 412 17.3.9.1 Examples of Contamination Incidents. . . . . . . . . . . . . . . . . . . . . 412 17.3.9.2 Radiological Considerations . . . . . 415 17.3.9.3 Important Nonradiological Considerations . . . . . . . . . . . . . . . . 415 17.3.9.4 Examples and Descriptions of Sources and Devices . . . . . . . . . . . . . . . . . . 415 17.3.10 Military . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 416 17.3.10.1 Examples of Contamination Incidents. . . . . . . . . . . . . . . . . . . . . 416 17.3.10.2 Radiological Considerations . . . . . 417 17.3.10.3 Important Nonradiological Considerations . . . . . . . . . . . . . . . . 419 17.3.10.4 Examples and Descriptions of Sources and Devices . . . . . . . . . . . 419 17.3.11 Decontamination and Decommissioning . . . . 419 17.3.11.1 Examples of Contamination Incidents. . . . . . . . . . . . . . . . . . . . . 421 17.3.11.2 Likely Radionuclides Involved . . . 421 17.3.11.3 Radiological Considerations . . . . . 421 17.3.11.4 Important Nonradiological Considerations . . . . . . . . . . . . . . . . 423 17.3.11.5 Examples and Descriptions of Sources and Devices . . . . . . . . . . . 423 17.3.12 Research . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 423 17.3.12.1 Examples of Contamination Incidents. . . . . . . . . . . . . . . . . . . . . 423 17.3.12.2 Radiological Considerations . . . . . 424 17.3.12.3 Important Nonradiological Considerations . . . . . . . . . . . . . . . . 426

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17.3.12.4

17.4

Examples and Descriptions of Sources and Devices . . . . . . . . . . . .426 17.3.13 Field Activities . . . . . . . . . . . . . . . . . . . . . . . . .426 17.3.13.1 Examples of Contamination Incidents . . . . . . . . . . . . . . . . . . . . .427 17.3.13.2 Radiological Considerations . . . . . .428 17.3.13.3 Important Nonradiological Considerations . . . . . . . . . . . . . . . .428 17.3.13.4 Examples and Descriptions of Sources and Devices . . . . . . . . . . . .430 17.3.14 Spacecraft and Space-Based Applications . . . .430 17.3.14.1 Examples of Contamination Incidents . . . . . . . . . . . . . . . . . . . . .431 17.3.14.2 Radiological Considerations . . . . . .431 17.3.14.3 Important Nonradiological Considerations . . . . . . . . . . . . . . . .435 Conclusions and Summary . . . . . . . . . . . . . . . . . . . . . . .435

18. Roles and Responsibilities of Responders to Radionuclide Contamination Incidents . . . . . . . . . . . . . . . . . . . . . . . . . . . .437 18.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .437 18.2 Small-Scale versus Large-Scale Incidents . . . . . . . . . . .438 18.2.1 Small-Scale Incidents . . . . . . . . . . . . . . . . . . . .438 18.2.2 Large-Scale Incidents . . . . . . . . . . . . . . . . . . . .438 18.3 Radionuclide Control Areas: Roles and Responsibilities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .439 18.3.1 Inner Contaminated Area . . . . . . . . . . . . . . . .442 18.3.1.1 Law Enforcement . . . . . . . . . . . . . .443 18.3.1.2 Emergency Medical Responders . .443 18.3.1.3 Radiation Safety/Health Physics . .443 18.3.1.4 Public Health . . . . . . . . . . . . . . . . .444 18.3.1.5 Other Emergency Responders . . . .444 18.3.2 Outer Contaminated Area . . . . . . . . . . . . . . . .444 18.3.2.1 Law Enforcement . . . . . . . . . . . . . .445 18.3.2.2 Emergency Medical Responders . .445 18.3.2.3 Radiation Safety/Health Physics . .446 18.3.2.4 Public Health . . . . . . . . . . . . . . . . .447 18.3.2.5 Other Emergency Responders . . . .448 18.3.3 Secured Area . . . . . . . . . . . . . . . . . . . . . . . . . . .448 18.3.3.1 Triage Area . . . . . . . . . . . . . . . . . . .448 18.3.3.2 Medical-Response Base . . . . . . . . .449 18.3.3.3 Decontamination Area . . . . . . . . . .449 18.3.4 Secured Area Perimeter and Control Point . . .450 18.3.4.1 Law Enforcement . . . . . . . . . . . . . .450 18.3.4.2 Emergency Medical Responders . .450

CONTENTS

20.5 20.6 20.7 20.8 20.9 20.10 20.11 20.12 20.13 20.14 20.15 20.16 20.17 20.18 20.19 20.20 20.21 20.22 20.23 20.24 20.25

/ xix

Cerium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .522 Cesium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .536 Cobalt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .552 Curium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .565 Europium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .575 Hydrogen (Tritium) . . . . . . . . . . . . . . . . . . . . . . . . . . . . .585 Iodine . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .594 Iridium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .605 Palladium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .617 Phosphorus . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .623 Plutonium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .632 Polonium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .657 Radium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .673 Rhenium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .686 Ruthenium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .692 Samarium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .707 Strontium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .715 Technetium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .733 Thorium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .739 Uranium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .754 Yttrium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .772

21. Dose-Assessment Methodologies . . . . . . . . . . . . . . . . . . . . .782 21.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .782 21.2 Collection of Data from the Individual . . . . . . . . . . . . . .783 21.3 Interpretation of External Contamination Monitoring Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .785 21.4 Interpretation of Monitoring Data in Cases of Internal Contamination . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .786 21.4.1 Methods for Making Rapid Decisions Based on Monitoring Results . . . . . . . . . . . . . . . . . . . . . .786 21.4.2 Use of Reference Values for Radionuclide Retention and Excretion . . . . . . . . . . . . . . . . . .787 21.4.3 Considerations Regarding the Time of Intake 792 21.4.4 Intake Estimates from Multiple Bioassay Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .792 21.4.5 Particle Size/Chemical Composition . . . . . . . .794 21.4.6 Intake Pathways . . . . . . . . . . . . . . . . . . . . . . . .794 21.4.7 Model and Data Uncertainty . . . . . . . . . . . . . .795 21.4.8 Modifications to the Reference Biokinetic Models . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .797 21.4.9 Direct Methods of Tritium Calculation . . . . . .799 21.4.10 Interpretation of Air Monitoring Data . . . . . . .800 21.5 Initial Assessment of External Dose and Internal Radionuclide Deposition . . . . . . . . . . . . . . . . . . . . . . . . .801

xx / CONTENTS 22. Research and Development . . . . . . . . . . . . . . . . . . . . . . . . 802 22.1 Decontamination Facilities for Removal of External Contamination . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 802 22.2 Instrumentation and Modeling for Assessment of Internal Contamination . . . . . . . . . . . . . . . . . . . . . . . . . 802 22.3 Bioassay Facilities and Automation . . . . . . . . . . . . . . . 804 22.4 Biomarkers and Devices for Biodosimetry . . . . . . . . . . 805 22.5 Software for Rapid Estimates of Organ Equivalent Dose and Effective Dose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 805 22.6 Decorporation Agents for Use Postexposure . . . . . . . . . 806 22.7 Medical Follow-Up of Exposed Populations . . . . . . . . . 808 22.8 Educational Programs . . . . . . . . . . . . . . . . . . . . . . . . . . 809 Appendix A. Radiological Recordkeeping Following an Incident . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.1 Records for Exposed Persons . . . . . . . . . . . . . . . . . . . . . A.2 Worker Records (Medical and Emergency Responders) A.3 Example Record Forms . . . . . . . . . . . . . . . . . . . . . . . . . Appendix B. Training . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B.1 Training Topics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B.1.1 Law-Enforcement and Emergency-Response Personnel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B.1.2 Emergency Medical Professionals . . . . . . . . . B.1.3 Radiation Workers . . . . . . . . . . . . . . . . . . . . . . B.1.4 Radiation-Safety Personnel . . . . . . . . . . . . . . . B.1.5 Public-Health Officials . . . . . . . . . . . . . . . . . . B.2 Drills and “Hands-On” Training . . . . . . . . . . . . . . . . . . B.3 Crisis and Risk Communications Training . . . . . . . . . .

810 810 811 812 819 820 820 821 822 822 823 824 824

Appendix C. Emergency-Responders’ Guidance on Radiation Risks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 825 C.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 825 C.2 Routes of Exposure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 826 C.2.1 Skin Contamination . . . . . . . . . . . . . . . . . . . . . 826 C.2.2 Inhalation and Ingestion . . . . . . . . . . . . . . . . . 826 C.3 Exposure to High Levels of Radiation . . . . . . . . . . . . . . 826 C.3.1 Early Health Effects . . . . . . . . . . . . . . . . . . . . 826 C.3.2 Late Health Effects . . . . . . . . . . . . . . . . . . . . . 827 C.4 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 828 Appendix D. Communications with the Media and the Public via the Media . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 829 D.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 829 D.2 Audiences to Consider . . . . . . . . . . . . . . . . . . . . . . . . . . 830 D.2.1 The Public . . . . . . . . . . . . . . . . . . . . . . . . . . . . 831

CONTENTS

D.3 D.4 D.5 D.6

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D.2.2 Medical Personnel . . . . . . . . . . . . . . . . . . . . . . .832 D.2.3 Off-Duty Emergency Responders . . . . . . . . . . .834 D.2.4 Those Outside the Affected Area . . . . . . . . . . .835 Spokesperson . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .835 Timeliness . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .837 Correcting Information Already Provided . . . . . . . . . . .838 Prescripted Statements to the Media . . . . . . . . . . . . . . .839 D.6.1 Brief Example . . . . . . . . . . . . . . . . . . . . . . . . . .839 D.6.2 Other Information to Convey . . . . . . . . . . . . . .840

Appendix E. Communicating with Patients and the Families of Patients Contaminated with Radionuclides . . . . . . . .842 Appendix F. Tables of Reference Values for Bioassay . . . . . .845 F.1 Actinium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .846 F.2 Americium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .847 F.3 Californium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .849 F.4 Cerium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .851 F.5 Cesium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .853 F.6 Cobalt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .856 F.7 Curium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .859 F.8 Europium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .860 F.9 Hydrogen (Tritium) . . . . . . . . . . . . . . . . . . . . . . . . . . . . .862 F.10 Iodine . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .863 F.11 Iridium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .865 F.12 Palladium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .868 F.13 Phosphorus . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .869 F.14 Plutonium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .871 F.15 Polonium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .876 F.16 Radium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .878 F.17 Rhenium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .879 F.18 Ruthenium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .880 F.19 Samarium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .884 F. 20 Strontium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .886 F.21 Thorium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .889 F.22 Uranium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .891 F.23 Yttrium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .893 Appendix G. Information Resources . . . . . . . . . . . . . . . . . . . . .894 Appendix H. Additional Radionuclide Exposure Case Studies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .896 H.1 Hanford Americium Accident (1976) . . . . . . . . . . . . . . .896 H.2 Chernobyl Nuclear Reactor Accident (1986) . . . . . . . . .902 H.2.1 Initial Response to the Accident . . . . . . . . . . .902 H.2.2 Classification of Acute Radiation Sickness . . .903

xviii / CONTENTS 18.3.4.3 18.3.4.4 18.3.4.5

Radiation Safety/Health Physics . 450 Public Health . . . . . . . . . . . . . . . . . 451 Other Emergency Responders . . . 451

19. Instrumentation to Measure Radionuclide Contamination . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 452 19.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 452 19.2 Direct (in vivo) Measurements of Body or Organ Content . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 454 19.2.1 Fixed-Facility Whole-Body Counters . . . . . . . 454 19.2.2 Fixed-Facility Special Monitoring Systems . . 456 19.2.3 Transportable Whole-Body Counters . . . . . . 458 19.2.4 Survey Meters for Large-Population Screening Measurements . . . . . . . . . . . . . . . . . . . . . . . . . 458 19.2.5 Special Considerations for Children and Pregnant Women . . . . . . . . . . . . . . . . . . . . . . . 459 19.2.6 Skin-Contamination Monitoring . . . . . . . . . . . 459 19.2.7 Wound Monitoring . . . . . . . . . . . . . . . . . . . . . . 461 19.2.7.1 Wound Monitors. . . . . . . . . . . . . . . 461 19.2.7.2 Survey Meters . . . . . . . . . . . . . . . . 464 19.2.7.3 Sequential Measurements. . . . . . . 464 19.3 Indirect (in vitro) Measurements of Body or Organ Content . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 465 19.3.1 Urine Samples . . . . . . . . . . . . . . . . . . . . . . . . . 465 19.3.2 Fecal Samples . . . . . . . . . . . . . . . . . . . . . . . . . 466 19.3.3 Breath Samples . . . . . . . . . . . . . . . . . . . . . . . . 467 19.3.4 Blood Samples . . . . . . . . . . . . . . . . . . . . . . . . . 467 19.3.5 Nose Blows and Nasal Swabs . . . . . . . . . . . . . 467 19.3.6 Other Biological Samples . . . . . . . . . . . . . . . . 468 19.3.7 Analytical Methods and Techniques . . . . . . . . 468 19.4 Contamination Survey Instrumentation and Measurements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 469 19.4.1 Surface-Contamination Measurements . . . . . 473 19.4.1.1 Alpha-Emitter Monitoring . . . . . . 476 19.4.1.2 Beta- and Gamma-Emitter Monitoring . . . . . . . . . . . . . . . . . . . 476 19.4.1.3 Surface Monitoring with Swipes. . 477 19.4.2 Air Sampling . . . . . . . . . . . . . . . . . . . . . . . . . . 477 20. Dosimetry and Case Studies for Selected Radionuclides . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20.2 Actinium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20.3 Americium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20.4 Californium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

482 482 485 493 509

xxii / CONTENTS

H.3 H.4

H.2.3 Early Clinical Response and Treatment . . . . H.2.4 Bone-Marrow Transplantation . . . . . . . . . . . . H.2.5 Evacuated Population . . . . . . . . . . . . . . . . . . . H.2.6 Late Effects of Radiation . . . . . . . . . . . . . . . . . Goiânia Incident (1987) . . . . . . . . . . . . . . . . . . . . . . . . . National Institutes of Health and Massachusetts Institute of Technology 32P Incidents (1995) . . . . . . . . . H.4.1 NIH 32P Contamination . . . . . . . . . . . . . . . . . . H.4.2 MIT 32P Incident . . . . . . . . . . . . . . . . . . . . . . .

905 907 907 908 908 915 916 916

Appendix I. Validation and Verification of Calculational Approach . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 918 Appendix J. Pregnancy Categories for Drug Use . . . . . . . . . 920 Glossary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 921 Abbreviations and Acronyms . . . . . . . . . . . . . . . . . . . . . . . . . . . 943 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 946 The NCRP . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 993 NCRP Publications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1002 Index . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1013

1. Executive Summary The purpose of this Report is to provide guidance to those who may be called to respond to radionuclide contamination incidents. Such incidents may range from situations in which one or a few persons have received minor contamination while working in research, medical facilities, or industry to those in which large numbers of people are contaminated as a result of accidental or deliberate releases of large quantities of radionuclides. The focus of this Report is on the medical management of individuals exposed to and potentially contaminated with radionuclides in such incidents. Thus, it is directed to persons who would provide medical care and those who would perform radiation-safety functions. This Report is intended as an update and expansion of the National Council on Radiation Protection and Measurements (NCRP) Report No. 65, Management of Persons Accidentally Contaminated with Radionuclides (NCRP, 1980). The primary objective in the management of persons contaminated with radionuclides is to reduce the risk of health effects occurring either early after the contamination incident or later in life. Two types of effects are of concern. One type of effect is harmful tissue reactions, termed deterministic effects. Doses large enough (threshold doses) to damage a critical population of cells can result in serious tissue or organ malfunction and possibly death. These effects can occur early after the contamination incident or later in life. A second type is termed stochastic effects. Radiation damage to cellular deoxyribonucleic acid (DNA) can lead to the expression of cancer later in life and to heritable effects in offspring. Radiationinduced cancer has been observed to occur in humans as well as in experimental animals. Although they have not been observed in humans, the potential for radiation-induced heritable effects is a concern because they have occurred in experimental animals. Heritable effects and cancer are not unique to radiation. According to the American Cancer Society (ACS, 2008), about one in three persons will be diagnosed with cancer in their lifetimes and one in four persons will die of cancer. The risk of radiation-induced lethal cancer is relatively low, ~5.5 % Sv–1 (100 rem) (ICRP, 2007). The total effective dose for persons in the United States in 2006 was estimated to be 6.2 mSv (620 mrem) (NCRP, 2009). In a population of >105,000 survivors of the atomic bombs in Japan, after 50 y, only 1

2 / 1. EXECUTIVE SUMMARY 853 of nearly 17,000 cancers were attributable to radiation exposures which ranged from 0.005 to 4 Gy (0.5 to 400 rad) (Preston et al., 2007). This Report is intended to provide guidance and recommendations to medical and radiation-safety personnel for reducing external and internal radionuclide contamination levels and thus the risks of radiation-induced health effects. The risk of deterministic effects can be reduced by controlling radiation doses to below their threshold doses. The risk of radiation-induced stochastic effects can be reduced, but not necessarily eliminated, by minimizing the dose. Radiation doses can be controlled or minimized by removing external radionuclide contamination and treating with blocking agents to decrease internal depositions and decorporation agents to increase the rate of elimination of radionuclides from the body. This Report comprises two volumes, a Handbook and the Scientific and Technical Bases. The Handbook contains information for immediate “in-the-field” application to radiation contamination incidents. It is organized into four parts. Three sections in Part A provide Quick Reference Information for incident responders based on supporting information given elsewhere in the Report. Three sections in Part B contain essential information on medical and radiation-safety activities to be conducted at the site of a radionuclide contamination incident and prior to arrival at a hospital. This is followed by four sections in Part C that describe medical and radiation safety activities at the hospital. Part D comprises two sections providing guidance on medical follow-up of exposed persons and on handling contaminated decedents. Parts A, B, C and D are color coded for easy access by users. Concluding the Handbook is guidance on contamination control in medical facilities. Part E, the Scientific and Technical Bases volume is organized into seven sections and several appendices. It provides detailed scientific and technical information in support of the guidance in the Handbook. While much of the information in the Scientific and Technical Bases can be found through diligent examinations of other publications, few organizations involved in responding to radionuclide contamination incidents will have access to all these publications or the staff to research them. Therefore, the Scientific and Technical Bases are presented as a resource, supporting and supplementing the material in the Handbook and for possible use in training emergency-response personnel. Experience has shown that it would be extremely rare for trained radiation-safety and medical personnel to be first on the scene of any radionuclide contamination incident, whether it is an accidental spill in a laboratory or the deliberate release of a large

1. EXECUTIVE SUMMARY

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quantity of activity in a public place. It is more likely that the first on the scene will be a colleague in a work place and fire-protection and law-enforcement personnel in a public place. Also, contamination incidents can occur in locations where radiation-safety and medical personnel are not readily available. Therefore, this Report is intended for the broad spectrum of persons who may respond to radionuclide contamination incidents, those with limited knowledge of radiation-safety and medical response as well as the professionals with extensive knowledge. Handbook, Part A, is intended to provide quick reference information that may be useful to anyone responding to a radionuclide contamination incident regardless of their radiation knowledge and experience. It begins with Section 3, a Compendium of Radiation Facts and Guidance. To help ensure that all responders are communicating with clarity and understanding, a number of terms are defined as they apply to this Report. Also, relevant properties of radiations emitted by radionuclides are reviewed. Since medical and radiation-safety personnel should be working as a team in all radiation incidents, their individual responsibilities are briefly summarized as well as guidance on working onsite and making preliminary health and radiation assessments. Section 3 also summarizes guidance on the management of injured and contaminated persons, identifying priorities and the actions to be taken in assessing exposure situations, and providing appropriate medical responses. A decision tree (Figure 3.1) for managing persons contaminated with radionuclides is included to guide radiation safety and medical personnel. Section 4, Radiation-Safety Guidance for First Responders, is a quick review of the objectives for the first medical and radiationsafety personnel on the scene of a radionuclide contamination incident. These include providing medical aid to injured individuals, identifying irradiated and contaminated persons, detecting and identifying radioactive material, identifying sources of external radiation, controlling the contamination, and initiating decontamination of individuals and the site. All of this is to be achieved with attention to protection of exposed persons and the responders. If the incident is accompanied by serious hazardous situations like fire or explosive destruction of buildings and vehicles, evacuation of people may be the first priority. Guidance is provided in establishing radiation contamination control areas (e.g., an inner area containing the radionuclide contamination incident site; an outer contaminated area; and outside of these, an uncontaminated secured area where onsite medical and radiation assessments and decontamination of people should be performed). Guidance is also

4 / 1. EXECUTIVE SUMMARY provided to ensure protection of responding personnel. This is further discussed in Appendix C, Emergency-Responders’ Guidance on Radiation Risks. Section 5, Performing Surveys and Controlling Personnel and Area Contamination, provides a review of how to survey individuals, equipment and other surfaces. This section is supported by Section 19, Instrumentation to Measure Radioactive Contamination, in the Scientific and Technical Bases, which provides more detailed information on radiation detecting instruments. Section 5 provides a quick review of the instrumentation to be used for a particular survey and how it is to be used (e.g., the distance to hold the detector from the object or person being surveyed, calibrating the instrument, and correcting for background). Also important to first responders is guidance on personal protection and the equipment that should be used and how to use it (e.g., inspection of equipment for damage such as tears in gloves, taping at the wrists and ankles, where to wear dosimeters, when respiratory protection is advised, and also how to remove the equipment to maintain control of the contamination). The medical management activities described in Parts B, C and D are organized into nine stages as shown in Figure 3.1. The applicability of one or up to all nine stages depends upon the nature and consequences of the contamination incident. For example, a minor incident in a laboratory may be quickly assessed in Stage 1 as having no health consequences and the exposed individuals can be released to their home or workplace. On the other hand, a major incident involving a large release and intake of radionuclides, whether exposing one individual or many, could result in severe health consequences and even death. In such cases, all nine stages might be applicable. A description of the objectives and the recommended actions to be taken in each stage is the major focus of this two-volume Report. Handbook, Part B, Onsite and Prehospital Actions, addresses actions to be taken onsite to assess and control both radiation and medical aspects of a radionuclide contamination incident, Stages 1 through 3 in Figure 3.1. Stage 1, Medical Assessment (Section 6) describes the initial assessment of a contamination incident and of the persons exposed to radiation and/or radioactive materials. This normally occurs onsite in a triage area within a secured area as shown in Figure 4.1. Ideally this will be conducted by both medical and radiation-safety personnel (Section 18.3.3). The highest priority is to provide immediate emergency care to individuals who have been seriously injured. The next priority is to identify those who have been exposed

1. EXECUTIVE SUMMARY

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and contaminated and those showing psychological distress. If there is evidence of radioiodine intakes, consideration should be given to administer potassium iodide (KI) and if intakes of transuranics such as plutonium are suspected, treatment with diethylenetriaminepentaacidic acid (DTPA) should be considered, since in both cases, prompt treatment can be most effective. Stage 2, External Contamination Assessment (Section 7) provides a description of the steps to be performed onsite in the secured uncontaminated area by radiation-safety personnel, assisted by medical personnel as needed (Section 18.3.3). The objectives are to assess the contamination, examine for burns, wounds, shrapnel, and hot particles and make treatment and decontamination recommendations. Stage 3, External Decontamination (Section 8) by radiationsafety and medical personnel should take place onsite in a location specifically identified for that purpose (Section 18.3.3). It could be a shower room if the incident occurred in a work location. If the incident occurred in a public area, it could be in a nearby gym, temporary tented facility, or hospital. The objective is to control external contamination to avoid inadvertent intakes through skin or by inhalation or ingestion. A second objective of removing contamination is to reduce radiation doses to skin and the risk of dermal injuries and to decrease amounts of radionuclides in wounds and their possible absorption into blood. Handbook, Part C addresses activities that would normally occur away from the site of the contamination incident, in a clinic or hospital emergency department. Stage 4, Patient Evaluation and Emergency Care (Section 9). In most cases, the patients will have been assessed for external contamination and decontaminated, but some of those severely injured may have arrived without having been decontaminated. Thus, proper radiation protection practices should be enforced to prevent contamination of the hospital. The emergency department physicians should have in hand all of the available documentation about the incident and any preliminary dose-assessment information for the patient, including the potential for whole-body external exposures. The objective is to evaluate and provide emergency treatment for injuries and examine for possible whole-body radiation exposures. Unless administered onsite, consideration should be given to administering KI if there is evidence of radioiodine intakes and DTPA if intakes of transuranics such as plutonium are suspected since in both cases, prompt treatment can be most effective. Stage 5, Internal Contamination Assessment (Section 10) also occurs at an emergency facility or hospital. The objectives are to

6 / 1. EXECUTIVE SUMMARY verify the contamination and to evaluate the intake and radiation dose by determining the routes of intake, identifying the radionuclides and assessing their quantities with appropriate bioassay procedures. Information obtained during the previous stages will help ascertain how this is to be done. Specific information about the incident, air samples taken, location of the individual relative to a release of activity, and the length of time in a contaminated area can be very useful in establishing the routes of intake. Stage 6, Clinical Decision Guidance (Section 11), would usually take place at hospitals and involves analysis of the internal dose assessments to determine whether consideration should be given to decorporation therapy. Information on radiation dose, excretion, and nasal swabs are compared to model predictions to assess whether intakes of radionuclides exceed the Clinical Decision Guide (CDG) for the particular radionuclide. The objective is to reduce the risk of stochastic effects, cancer, to a level consistent with current regulatory guidance for responding to emergency situations and to prevent the risks of deterministic effects. To guide physicians in considering the need for medical treatment to achieve this objective, a new operational quantity, CDG is introduced. The numerical values of dose used as a basis for computing the CDG intake values for different radionuclides, excluding isotopes of iodine, in adults are: • 0.25 Sv (25 rem) (50 y effective dose) for consideration of stochastic effects [this represents about a 1.3 % lifetime risk of fatal cancer attributable to the radiation dose (ICRP, 2007)]; • 30 d RBE-weighted absorbed-dose value of 0.25 Gy-Eq (25 rad-Eq) for consideration of deterministic effects to bone marrow; and • 30 d RBE-weighted absorbed-dose value of 1 Gy-Eq (100 rad-Eq) for consideration of deterministic effects to the lungs. For radionuclides other than isotopes of iodine, the CDGs for children (0 to 18 y of age) and pregnant women are defined as one-fifth the adult value. CDG values for 131I are based on the U.S. Food and Drug Administration (FDA) recommendations (FDA, 2001) that KI be administered to adults >40 y of age if the projected dose to thyroid is ≥5 Gy (500 rad), to adults 18 to 40 y of age if the projected dose is ≥0.1 Gy (10 rad), and to pregnant or lactating women or persons <18 y of age if the projected dose is ≥0.05 Gy (5 rad). CDGs are tabulated for 25 radionuclides. Stage 7, Medical Management (Section 12), provides guidance on treatment that will most often occur in hospitals, at least initially.

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The goal is to select a treatment modality that will enhance the excretion of the particular radionuclide from the body with minimal impact upon the health of the patient and thus reduce the lifetime risk of cancer and reduce the potential for both deterministic and stochastic effects. Treatment would include appropriate clinical follow-up. Handbook Part D addresses actions that would occur after contaminated persons are discharged from hospital or in the event of death of contaminated persons. Stage 8, Follow-Up Medical Care (Section 13), should involve a long-term plan for patients who experience radionuclide contamination sufficient to be considered for treatment as well as those who received treatment. They should be admitted to a registry for maintaining records of radiation dose, treatment, surveillance for subsequent malignancy, and other effects, including psychosocial. Stage 9, Contaminated Decedents (Section 14) provides guidance to protect medical examiners and mortuary personnel from excessive radiation exposure and to ensure that burial or cremation of the body is conducted with care to control radionuclide contamination. The Handbook concludes with Section 15, Contamination Control in Medical Facilities. Few emergency departments are staffed and equipped to handle radionuclide contaminated patients. Even in hospitals with radiation medicine departments there may not be sufficient expertise to assist emergency department personnel in controlling the spread of radionuclide contamination. While normal hospital sterility practices will in most cases be adequate, radionuclides do present special control problems with respect to the patients, the staff, and the facilities. However, treatment of contaminated patients is first priority. Guidance in this section is directed to those control issues. The Scientific and Technical Bases volume, Part E, comprises seven sections and 10 appendices. Although much, but not all, of the information can be found in other publications, it is compiled here to help explain some of the recommendations in the Handbook and offer supplemental information. Section 16, Overview of Radiobiology Concepts Pertinent to Radionuclides, is included in recognition of the fact that irradiation by radionuclides released in a contamination incident differs in major ways from irradiation by external sources. For example, radionuclides deposited upon the skin or in the body usually irradiate only part of the body, compared with external radiation that can irradiate the whole body (exceptions are when external beams are partially shielded or when radiation for therapeutic purposes is directed to specific tissues and organs). Also, except for

8 / 1. EXECUTIVE SUMMARY radionuclides with very short half-lives, radiation from radionuclides is delivered over a period of time that can range up to years. Radiation from external sources is delivered over a relatively short period of time, depending upon how long the individual is in the beam and how long a radiation generator is operating (exceptions are therapy doses protracted over weeks or months and natural background). These exposure differences can result in significantly different health consequences. When a person is in the vicinity of an uncontrolled radionuclide source, contamination of skin is more likely than an intake by ingestion or inhalation. Ingestion requires the source to be in food or water or moved to the mouth area by physical contact, and inhalation requires the source to be airborne in a person’s breathing space. Section 17, Settings in Which Individuals May be Contaminated with Radionuclides, describes numerous scenarios in which radionuclides can be released, inadvertently or deliberately. The radionuclides of concern are identified for each scenario. Also tabulated are deep-dose-rate and skin-dose-rate values from external sources. Contamination incidents can be small-scale such as an accidental release in a laboratory or large-scale, such as a rupture of a large radioactive source in a public place. Nearly all are accidental, but deliberate releases are becoming of greater concern with the increase of unaccounted-for (orphan) radionuclide sources. Section 18, Roles and Responsibilities of Responders to Contamination Incidents, supplements the information given in Sections 3, 4, 5, 6, 7 and 8 for onsite and prehospital actions. It provides additional guidance on designating controlled areas and defining responsibilities for potential responders of various disciplines such as law-enforcement, fire-protection, medical, radiation-safety, and public-health personnel. Specific responsibilities are described for radiation-safety personnel ranging from determining the number of trained radiation-safety personnel required at the scene and conducting the necessary monitoring to documenting the actions for the incident commander. Section 19, Instrumentation to Measure Radioactive Contamination, supports Sections 3, 5, 6, 7, 8, 9 and 10 with detailed descriptions of the instrumentation and the analytical procedures required to detect radionuclide contamination and assess external and internal contamination. This includes initial qualitative surveys to identify the presence of radioactive material and relatively precise quantitative measurements for the purpose of estimating radiation doses. Section 20, Dosimetry and Case Studies for Selected Radionuclides, provides biokinetic and dosimetric information for the

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radionuclides that are identified in Section 17 as being potentially involved in radionuclide-contamination incidents. This section is also the source of the dosimetric data tabulated in the Handbook. The information presented is intended primarily for assessment of radiation dose during the early hours or days after exposure. Forms of radionuclides and modes of intake were selected in an attempt to provide typical estimates of tissue dose from types of exposure that seem most likely based on case studies. Relevant case studies are included with each element covered in Section 20. It is assumed that consistent with past experience, while skin contamination is more likely, nearly all future inadvertent intakes of radionuclides will be by the inhalation route, although intakes by ingestion and through wounds or skin absorption are not excluded. Readers particularly interested in issues related to the behavior of radionuclides in wounds and their treatment should also consult NCRP Report No. 156, Development of a Biokinetic Model for Radionuclide-Contaminated Wounds and Procedures for Their Assessment, Dosimetry and Treatment (NCRP, 2006a). Section 21, Dose-Assessment Methodologies, provides an overview of information and considerations for evaluating intakes and radiation doses needed to make treatment decisions. Although much information can be collected about an incident, the most important is the information collected from the contaminated individual. This section illustrates the use of data obtained using the instrumentation and procedures described in Section 19, to evaluate external and internal doses to exposed individuals. Section 22, Research and Development, describes current and needed research and development to improve the effectiveness of responding to radionuclide contamination incidents. A concern is the paucity of facilities for decontaminating individuals, especially large numbers of people, and assessment of external and internal contamination including in vivo and in vitro bioassay capabilities. Cooperative agreements between local emergency planners and national laboratories and nuclear power plants around the country should be encouraged where they do not already exist. Research needs include improved instrumentation and modeling capabilities for internal dose assessment, development of biomarkers for biodosimetry, improved dosimetry software, and decorporation drugs. Current research on treating for internal contamination, described in this section, shows promise and should be continued. Appendices, glossary, and references complete the Scientific and Technical Bases. Ten appendices provide specific information on several important topics relevant to both the Handbook and the Scientific and Technical Bases and are referenced in the appropriate

10 / 1. EXECUTIVE SUMMARY sections of both volumes. Topics covered in the appendices include record keeping, training, guidance on radiation risks for emergency responders, and communicating with the media, members of the public, and with patients and their families. Reference bioassay tables are included as an appendix for use in assessment of internal contamination (Sections 10 and 21) and another appendix contains several case studies as examples of radiation-safety and medical responses to major radionuclide exposure incidents. A list of useful resource documents is included as another appendix. This is followed by an appendix that provides details on validation and verification of the models, software, and associated parameters used for the biokinetic and dosimetric calculations in this Report and a final appendix that lists categories for drug use in pregnancy.

2. Introduction In 1980, NCRP issued Report No. 65, Management of Persons Accidentally Contaminated with Radionuclides (NCRP, 1980). The report was a manual intended primarily for use by physicians in the management of accidental radionuclide contamination incidents. This 1980 report included a quick-reference guide describing steps to be taken in providing medical care and radiation protection to contaminated individuals. This guidance was supported by information on assessing levels of contamination of individuals, the potential health risks to be considered in decisions concerning treatment, and approaches to decorporation and medical intervention. Report No. 65 was mostly directed to accidental contamination incidents that might occur in hospitals, research institutions, nuclear power plants, national laboratories, military installations, and nuclear-weapons facilities. Its value is indicated by the fact it was reprinted several times and over 18,000 copies have been distributed worldwide. While the general guidance provided in Report No. 65 remains valid, the practice of radiation safety and the tools used by radiationsafety practitioners and the medical profession have progressed considerably since the report was published. Further, in addition to occupational contamination accidents, several contamination incidents have exposed members of the public and with the expanding use of radiopharmaceuticals there is an increasing probability of treatment errors resulting in overdoses of radionuclides. Therefore, NCRP chose to update Report No. 65 with the present Report to reflect improved medical and radiation protection technology and the experience gained in responding to incidents involving radionuclide contamination of multiple individuals including members of the public. Furthermore, there is significant concern that the frequency and seriousness of public contamination incidents will increase due to increasing numbers of unaccounted-for (orphan) radionuclide sources throughout the world, as well as the threat of deliberate releases of radionuclides in terrorist activities. This Report addresses radiation exposures that are beyond the regulatory framework and thus are not a matter of regulatory compliance. While the emphasis of this Report is on the medical management of contaminated persons, guidance is also directed to important supporting radiation-safety functions, broadly termed 11

12 / 2. INTRODUCTION “health physics.” A major component of this guidance is the incorporation of the best available data on radionuclide biokinetics and dosimetry. Guidance on medical management, radionuclide contamination levels, use of instrumentation, and radiation-safety procedures is drawn from, and referenced to, appropriate peerreviewed publications and other credible documents including numerous publications from national and international bodies. The threat of deliberate releases of radioactive material is the subject of NCRP Report No. 138, Management of Terrorist Events Involving Radioactive Material (NCRP, 2001a). That report covers three areas in considerable depth. First, descriptions are given of types of nuclear terrorist incidents and the decision-making processes to be faced by law enforcement, medical staff, and publichealth officials regarding potential human casualties with their early and late health issues and psychological consequences. Second, detailed guidance is provided for management of a nuclear terrorist disaster including command and control responsibilities by local and national government agencies and their officials, communication requirements for emergency-response personnel, and communications to the media and public. Radiation protection considerations are also addressed. Third, recommendations for preparing for possible future nuclear terrorist activities are provided in detail. In another effort related to responses to possible terrorist activities, NCRP published Commentary No. 19, Key Elements of Preparing Emergency Responders for Nuclear and Radiological Terrorism (NCRP, 2005a). This commentary provides a technical basis for the support of preparedness activities such as the development of emergency-responder protocols, equipment procurement recommendations, and the frequency and content of training exercises. Report No. 138 and Commentary No. 19 are highly recommended as companions to this Report as are others listed in Table 2.1. 2.1 Purpose of this Report The primary purpose of this Report is to update the information and guidance originally given in NCRP Report No. 65 (NCRP, 1980) and broaden its scope to address the potential contamination not only of radiation workers, as was the focus of Report No. 65, but also members of the public, including children. Because many changes have occurred since 1980 in radiation protection concepts, philosophy and practice, much of the text in Report No. 65 required rewriting even though much is still valid technically. Information regarding specific radionuclides has been enhanced in this Report by updated data and summaries of biokinetic and dosimetric models

2.1 PURPOSE OF THIS REPORT

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TABLE 2.1—Examples of publications recommended for further information.a Development of an Extended Framework for Emergency Response Criteria

IAEA (2005a)

Disaster Preparedness for Radiology Professionals Response to Radiological Terrorism: A Primer for Radiologists, Radiation Oncologists and Medical Physicists

ACR (2006)

Generic Procedures for Medical Response During a Nuclear or Radiological Emergency

IAEA (2005b)

Guidebook for the Treatment of Accidental Internal Radionuclide Contamination of Workers

Bhattacharyya et al. (1992)

Handbook for Responding to a Radiological Dispersal Device, First Responder’s Guide–The First 12 Hours

CRCPD (2006)

Key Elements of Preparing Emergency Responders for Nuclear and Radiological Terrorism

NCRP (2005a)

Management of Terrorist Events Involving Radioactive Material

NCRP (2001a)

Manual for First Responders to a Radiological Emergency

IAEA (2006)

Medical Management of Radiological Casualties Handbook

AFRRI (2003)

Planning Guidance for Protection and Recovery Following Radiological Dispersal Device (RDD) and Improvised Nuclear Device (IND) Incidents

FEMA (2008)

Population Monitoring in Radiation Emergencies: A Guide for State and Local Public Health Planners

CDC (2007)

Procedures for Medical Emergencies Involving Radiation

HPS (2006)

Protecting People Against Radiation Exposure in the Event of a Radiological Attack

ICRP (2005a)

Radioactivity and Health: A History

Stannard (1988)

REMM: Radiation Event Medical Management

DHHS (2009)

aMore information on how to contact these organizations and obtain these reports and other related information is provided in Appendix G and in the reference list.

14 / 2. INTRODUCTION of the International Commission on Radiological Protection (ICRP). These models were used to calculate organ absorbed-doses, wholebody and organ retention values, and excretion rates of radionuclides. These recently developed models and improved knowledge of health risks as a consequence of the intake of radionuclides provide increased confidence in decisions concerning vital treatment. While still focused on the management of persons contaminated in occupational settings, this Report also addresses the potential for exposure of members of the public through accidents involving radionuclides released from laboratories, medical facilities, power plants, industries, and material transport. This Report can also provide guidance in cases of a medical incident involving the administration of a radiopharmaceutical to the wrong patient, the wrong radiopharmaceutical to a patient, an excessive activity of a radiopharmaceutical to a patient, or the leakage of a sealed source implanted in a patient for brachytherapy. Special emphasis is given to managing persons contaminated as a result of terrorist activities since these could involve large numbers of people. An aspect of such incidents is the potential for social, psychological and behavioral consequences to comprise possibly more significant concerns for health and safety than potential radiation injuries. While this topic was addressed in NCRP Report No. 138 (NCRP, 2001a), it is also brought to the attention of readers of this Report because the potential for such adverse consequences exists for all types of contamination incidents and, thus, is not limited to incidents associated with terrorist’s activities. This Report also supplements the guidance given in NCRP Commentary No. 19 (NCRP, 2005a). The primary objective in the management of persons contaminated with radionuclides is to reduce the risk of health effects occurring either early after the contamination incident or later in life. Two types of effects are of concern. One type of effect is harmful tissue reactions, termed deterministic effects. Doses large enough (threshold doses) to damage a critical population of cells can result in serious tissue or organ malfunction and possibly death. These effects can occur early after the contamination incident or later in life. The risk of deterministic effects can be reduced by controlling radiation doses to below their threshold doses. A second type is termed stochastic effects. Radiation damage to cellular DNA can lead to the expression of cancer later in life and to heritable effects in offspring. Radiation-induced cancer has been observed to occur in humans as well as in experimental animals. Although they have not been observed in humans, the potential for radiation-induced heritable effects is a concern because they have occurred in experimental animals. The risk of radiation-induced stochastic effects can be reduced, but not necessarily eliminated, by minimizing the

2.3 ORGANIZATION OF THIS REPORT

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radiation dose. Heritable effects and cancer are not unique to radiation. According to the American Cancer Society (ACS, 2008), about one in three persons will be diagnosed with cancer in their lifetime and one in four persons will die of cancer. Even after 50 y, among >105,000 Japanese survivors of the atomic bombs, only 853 of nearly 17,000 cancer cases were attributable to radiation (Preston et al., 2007). The objective of this Report is to provide guidance and recommendations to medical and radiation-safety personnel for reducing external and internal radionuclide contamination levels and thus the risks of radiation-induced health effects. 2.2 Target Audiences of this Report Since 1980, the varieties of situations within which persons are likely to become contaminated with radionuclides have changed and public concern has increased. Therefore, this Report is directed at an audience broader than that originally intended for Report No. 65. The audiences for the present Report now include essentially all emergency responders and not simply medical personnel. This Report is written specifically for those who might become directly involved in responding to incidents in which a person or persons may be contaminated with radionuclides. In an occupational setting, this could include coworkers, supervisors, managers, radiation-safety personnel (e.g., health physicists and radiation safety officers), nurses, and physicians. In an accidental release into the environment from a missing source, nuclear power plant incident, transportation accident, industrial application, medical facility, or military, aviation or space activity, an even broader spectrum of persons could be involved. Such persons include publichealth officials, emergency medical personnel, hospital staff, fireprotection personnel, law-enforcement officials, and local, state and federal government officials. In the event of a terrorist incident, first responders could include untrained citizens as well as radiation-safety, emergency medical, fire-protection, and lawenforcement personnel. It is also intended that this Report be a source of information to be consulted by public officials well in advance of any serious radionuclide contamination incident to forestall the potential for chaos and panic should such an incident occur and to provide information needed in training programs for all relevant responding personnel. 2.3 Organization of this Report For the convenience of users, this Report is in two volumes. The Handbook is intended as an update of NCRP Report No. 65 and,

16 / 2. INTRODUCTION like Report No. 65, is of a size that can be readily accessible. It contains information, recommendations and advice relevant to management of persons contaminated with radionuclides in a broad range of potential incidents. The Handbook is for “in-the-field” use in responding to radionuclide contamination incidents. The Scientific and Technical Bases volume supports the recommendations and advice in the Handbook. There is some redundancy throughout these two volumes. This was deliberate because it was realized that many users of this Report will read only those sections most relevant to their technical discipline or interest. NCRP Report No. 65 was mostly directed to accidental contamination incidents in locations where trained radiation professionals and physicians are onsite or readily available. The broad spectrum of potential future contamination incidents means that many first responders will not be so well trained or have access to information and the relevant scientific literature. Therefore, this Report comprises much basic information that will be familiar to most medical and radiation-safety professionals, but may be of value as a review for them and a useful indoctrination for those planners and responders not trained in radiation sciences. 2.3.1

Management of Persons Contaminated with Radionuclides: Handbook

The Handbook is in four parts. Part A comprises quick reference information that may be of use to all first responders to a contamination incident. The management of persons contaminated with radionuclides will usually occur at two locations, at the site of the contamination incident and at a hospital or other emergency facility. Part B describes activities most likely to occur onsite and prehospital. Part C describes activities that should occur away from the incident site such as in a hospital. Part D provides guidance on actions that would occur post-hospital. These four parts are color coded for quick access. The actions taken by medical and radiation-safety professionals at these three venues can be considered to occur in several stages. Patient flow through various stages in the management of contaminated persons has been described by Bushberg et al. (2007) and by the U.S. Department of Health and Human Services (DHHS, 2009). This Report defines nine stages and courses of actions to be taken to achieve a favorable outcome for exposed individuals. These nine stages are described in Table 2.2 with the objectives associated with each. Figure 3.1, decision tree for management of persons

TABLE 2.2—Stages in the management of contaminated persons. Stage

Location

Objectives

• Stabilize life-threatening problems (consider admission to emergency facility) • Treat other injuries and observe for psychological distress among exposed persons • Identify exposed persons and those externally and/or internally contaminated • Document incident

2. External Contamination Assessment (Section 7)

Onsite triage area (Figure 4.1)

• Locate contaminated body area, including orifices • Identify hot particles, shrapnel, and contaminated debris • Identify and quantify radionuclides • Evaluate potential for skin injury • Confirm internally-contaminated individuals • Assess need for decontamination • Document contamination on body surfaces

3. External Decontamination (Section 8)

Onsite decontamination area (Figure 4.1)

• Control external contamination • Reduce radiation dose to skin • Reduce amounts of radionuclides in wounds • Document

4. Patient Evaluation and Emergency Care (Section 9)

Hospital

• Evaluate and treat patients with injuries and psychological distress • Evaluate patients for evidence of radiation sickness • Evaluate patients with skin burns, contaminated wounds, and intake for emergency treatment • Confirm possible internal contamination • Document evaluations and treatment

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Onsite triage area (Figure 4.1)

2.3 ORGANIZATION OF THIS REPORT

1. Medical Assessment (Section 6)

Stage

Location

Objectives

5. Internal Contamination Assessment (Section 10)

Hospital

• Determine routes of entry into body • Identify radionuclides and their physical/chemical form • Determine radiation doses using in vivo and in vitro bioassay procedures • Document internal contamination and radiation doses

6. Clinical Decision Guidance (Section 11)

Hospital

• Evaluate radiation dose with respect to CDG • Provide guidance to physicians making treatment decisions • Document guidance

7. Medical Management (Section 12)

Hospital

• Begin appropriate decorporation therapy • Evaluate treatment efficacy • Clinical follow-up • Document treatment

8. Follow-Up Medical Care (Section 13)

Where appropriate

• Late deterministic effects • Latent acute radiation syndrome • Psychosocial indications • Internal contamination • Cancer • Epidemiology studies • Patient record documentation

9. Contaminated Decedents (Section 14)

Hospital and mortuary

• Protect medical examiners and mortuary professionals from radiation exposures • Control radionuclide contamination of individuals and facilities • Ensure proper disposal of decedent • Document

18 / 2. INTRODUCTION

TABLE 2.2—(continued).

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contaminated with radionuclides, depicts the normal flow of an exposed person through all stages, but there may be exceptions in which certain stages might be bypassed depending upon the status of the person involved. Both Figure 3.1 and Table 2.2 are intended as guides in the use of this Handbook. 2.3.2

Part A: Quick Reference Information

Part A begins with Section 3, which is largely a tabulation of radiation facts, definitions of terminology and specific guidance for both radiation-safety and medical personnel who are first on the scene. The information is tailored for easy use by early responders to contamination incidents and is referenced to other parts of the Report for details. For example, a foldout decision tree for managing persons contaminated with radionuclides is included. This chart is color coded to direct readers to appropriate sections of the Handbook for specific information. Section 4 offers guidance for radiation-safety and medical first responders to an incident. These responders will likely be familiar with guidance relevant to their professions, but not to the others’. Having this knowledge should enhance their ability to work as a team. Section 5 is a refresher on performing surveys and controlling contamination at the incident site. Again, this information will be familiar to health physicists and other radiation-safety professionals, but there may be occasions when persons not having this knowledge and experience will be on the scene. 2.3.3

Part B: Onsite and Prehospital Actions

Part B provides onsite guidance. The first three stages of managing contaminated persons will usually occur at the site of the incident, before individuals are transported from the area. The first stage is medical assessment in which potentially-exposed persons are screened for life-threatening problems and for evidence of radiation exposure and radionuclide contamination. Depending upon the nature of the contaminating incident, it may be that not all persons at the site will be exposed and contaminated. This assessment will usually occur in a designated triage area at the site of the incident. The highest priority action is stabilization of patients with life-threatening injuries. Preliminary screening for external and internal contamination is an important part of this assessment in that it may determine whether the exposed persons require further evaluation and/or evacuation. Medical assessment activities are described in Section 6.

20 / 2. INTRODUCTION The second stage is external contamination screening which also will occur onsite in a designated triage area. Guidance is given in Section 7 on locating and identifying contamination, evaluating possible skin injuries, and assessing needs for treatment and decontamination and possibilities of intakes. The third stage is external decontamination. This action should also be conducted in a secured controlled area, onsite or at a special facility when available. The steps and equipment necessary to externally decontaminate individuals are described in Section 8. 2.3.4

Part C: Patient Management at Hospital

Part C, addresses actions to be taken after the patient is transported to a hospital or other emergency medical facility (Table 2.2). The fourth stage is evaluation and emergency care of the patient by emergency department medical personnel. As described in Section 9, it includes evaluation and treatment of injuries, diagnosis of radiation exposure, and initial treatment decisions. The fifth stage is an assessment of internal contamination. Collection of biological samples such as nasal swabs and excreta should occur as soon as possible. If collected onsite, it should be done in a controlled area to prevent inadvertent contamination of the samples. Normally, collection of samples will occur at the emergency facility as would other bioassay procedures such as wholebody counting. Interpretation of bioassay data and assessment of contamination and radiation dose will usually occur after the exposed person has been transferred to a medical facility as described in Section 10. The sixth stage, an evaluation of the radionuclide intake and radiation exposure leading to decisions about decorporation treatment is described in Section 11. A new operational quantity, Clinical Decision Guide (CDG), is introduced in this section to assist physicians in making treatment decisions. The seventh stage is decorporation treatment. Drugs and treatment modalities are recommended for a broad range of radionuclides in Section 12. 2.3.5

Part D: Patient Management Post-Hospital

The eighth stage is medical follow-up of exposed persons that will vary depending upon many factors such as the nature of the contamination, decontaminating and decorporation treatment, physical and psychosocial trauma. Section 13 provides guidance for ensuring patients receive appropriate care after discharge from a medical facility.

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The ninth stage is the handling of contaminated deceased persons. Fatalities are a possible result of the more serious contamination incidents. Section 14 provides guidance on controlling contamination and protecting the professionals from unnecessary radiation exposures. Because the first priority is the care and management of contaminated persons, contamination of emergency department and other areas of the hospital can occur. Therefore, the Handbook concludes with Section 15, providing guidance in controlling the spread of contamination and decontaminating medical facilities and equipment. 2.3.6

Management of Persons Contaminated with Radionuclides: Scientific and Technical Bases

The second volume, Part E, supports the advice and recommendations in the Handbook with detailed information. It is more comprehensive than might be necessary for most radiation-trained professionals, but it is also intended to provide basic information for those potential responders to radionuclide-contamination incidents who have little knowledge of radiobiology and radiation protection and who do not have ready access to the scientific literature. These might include such professionals as industrial hygienists, nurses, physicians, law-enforcement and fire-protection personnel and emergency medical personnel. Section 16 is an overview of radiobiology concepts pertinent to the intake of radionuclides. This includes information on radiation, radionuclides, radiation dosimetry, and the human health risks of radiation exposure. Section 17 provides descriptions of known and potential contamination situations that could lead to small- and large-scale radionuclide contamination of workers and members of the public. It also identifies the radionuclides most likely to be encountered. Section 18 includes a comprehensive description of the roles and responsibilities of responders in designated control areas of contamination sites. Section 19 describes the technology and instrumentation available to detect and measure radionuclide contamination. The detection and measurement of radionuclides on and in humans have improved dramatically in the last 25 y. This advanced technology is described with recommendations for its use in specific situations. The chemical, biological and dosimetric properties of important radionuclides of 24 elements identified in Section 17 are addressed in Section 20. Extensive information is provided on the potential for exposure, methods of monitoring and measuring intakes, biokinetics

22 / 2. INTRODUCTION of specific compounds, dosimetric models, and estimates of radiation doses. Absorbed and effective doses per unit intake are tabulated for these radionuclides. Readers particularly interested in issues related to the behavior of radionuclides in wounds and their treatment should also consult NCRP Report No. 156 (NCRP, 2006a). NCRP devoted considerable attention to the quality assurance issues by having members carry out independent calculations and reviews of the tabulated results. Further information on validation and verification of the models and dose calculations in this Report can be found in Appendix I. Case studies are included in Section 20 for several radionuclides to illustrate how some contamination incidents have been managed in the past, demonstrate some important principles or contain unique lessons regarding radiation protection, and provide important biokinetic and dosimetric information gained from real contamination incidents. The information in this section is available elsewhere, but its compilation in this Report from several ICRP publications and numerous journal publications is a valuable resource well beyond the purpose of this Report. Difficult decisions are required in responding to radionuclide contamination incidents. One of the most difficult issues is the evaluation of the exposures. Since direct measurements are rarely possible, indirect methods are required. Section 21 describes the collection of data and its interpretation using various dose-assessment methodologies that are the bases for making these decisions. Finally, Section 22 reviews current research and needs for additional research and development, facilities, and education to better prepare for managing persons contaminated with radionuclides, in both small- and large-scale incidents. Since the publication of NCRP Report No. 65, little research has been directed towards the technical problems associated with management of significant radionuclide contamination of large numbers of persons. Although several new approaches are being explored, a serious need continues for more effective methods for removing radionuclides from the body (decorporation) and therapy for reducing the effects of radiation exposure. Recommendations for future research in these areas as well as in biodosimetry, instrumentation and software, and needs for planning for the management of potential incidents as well as for facilities to handle contaminated persons are detailed in Section 22. Ten appendices are included in this Report. Appendix A contains example forms that might be used to document information concerning contamination incidents, radiation dose assessments, and patient treatments. Appendix B gives guidance in training potential

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early responders, while Appendix C explains the potential radiation risks to those who respond to contamination incidents and perform needed services ranging from law-enforcement, radiation-safety personnel to medical professionals. Appendix D provides some guidelines in communicating with the media, and Appendix E, communicating with contaminated individuals and their families. Appendix F supports Section 21 by providing reference tables for interpreting bioassay data. Appendix G provides information on other useful resources, such as documents, books and websites dealing with topics covered in this Report. Appendix H describes the responses to several serious contamination incidents, illustrating the medical treatment of internally-deposited radionuclides as well as the ingenuity and resourcefulness that can be employed when an incident occurs remotely from radiation expertise and facilities. Validation and verification of the models, software, and associated parameters used for the biokinetic and dosimetric calculations in this Report are detailed in Appendix I and Appendix J lists categories for drug use in pregnancy. While NCRP has adopted SI units, this Report generally includes both SI units and units of the previous system (rad, rem and curie), which are still in use by many in the United States.

Part A: Quick Reference Information 3. Compendium of Radiation Facts and Guidance 3.1 Introduction This section comprises a compendium of information for the medical and radiation-safety personnel who would respond to incidents in which persons may be contaminated with radionuclides. It consists of information, facts and guidance that may be needed quickly by those who are first on the scene of a radionuclide release. The information is deliberately brief for ease of rapid access. The locations of more detailed information are indicated throughout this compendium. The medical and radiation-safety personnel using this Report will likely be familiar with the information relevant to their disciplines, but not necessarily with information relevant to other disciplines, which is important as they work as a team to manage persons contaminated with radionuclides. The information provided also may be helpful to other personnel who might be early responders to contamination incidents. Incidents that result in contamination of persons with radionuclides can range from small-scale where one or a few individuals are contaminated with small amounts of radionuclides to large-scale incidents where perhaps hundreds of individuals may be contaminated with large quantities of radionuclides. Such incidents can be inadvertent releases in laboratories, hospitals, power plants, industry, and the military or they may be deliberate releases as a result of terrorist activities. Further information can be found in Section 17. 3.1.1

Organizations Offering Radiological Incident Assistance

A number of federal and state organizations are available to offer radiological emergency-response assistance in the event of a 25

26 / 3. COMPENDIUM OF RADIATION FACTS AND GUIDANCE radiological or nuclear incident or emergency, at the request of the affected community or facility. Federal organizations include the following: • Armed Forces Radiobiology Research Institute (AFRRI) • U.S. Department of Energy (DOE) - field offices and national laboratories - Radiological Assistance Program - Radiation Emergency Assistance Center/Training Site (REAC/TS) • U.S. Department of Health and Human Services (DHHS) - Centers for Disease Control and Prevention (CDC) - U.S. Food and Drug Administration (FDA) - Radiation Event Medical Management (REMM) Guidance (DHHS, 2009) • U.S. Environmental Protection Agency • Federal Radiological Monitoring and Assistance Center Information on how to contact these and other organizations for assistance and to obtain copies of publications and other related information is provided in Appendix G. 3.1.2

Terminology

In the management of persons contaminated or potentially contaminated with radionuclides, communication among the radiation-safety and medical personnel and other responders is important. To minimize misunderstandings and confusion, the terminology used must have clear meanings. To eliminate some of the current ambiguity in the meaning and use of some of this terminology, the following important terms are defined as used throughout this Report. There are some disparities in how these terms are used among government agencies, NCRP and ICRP. The following are generally consistent with NCRP and, in most cases, ICRP usage. These terms also appear in the Glossary along with other terms relevant to the subject of this Report. • absorption: Movement of material to blood regardless of mechanism. Generally applies to the uptake into blood of soluble substances and material dissociated from particles. • absorption functions: Mathematical equations describing the rate of transfer of radionuclides into blood after deposition on skin, in wounds, in the gastrointestinal (GI) tract and in the respiratory tract (equations can be exponential, polynomial or constant relationships). • activity median aerodynamic diameter (AMAD): Median diameter of airborne radioactive particles having the same

3.1 INTRODUCTION

•

•

•

•

•

•

•

/ 27

aerodynamic properties as unit density spheres. Fifty percent of the activity (aerodynamically classified) in the aerosol is associated with particles greater than the AMAD. A lognormal distribution of particle sizes is assumed. bioassay: Any procedure used to determine the nature, activity, location or retention of radionuclides in the body by direct (in vivo) measurement or by indirect (in vitro) analysis of material excreted or otherwise removed from the body. clearance: The action that results in the movement of radioactive material from the site of deposition in tissues and organs. This action can be natural or induced by therapeutic means. - pathways: Routes by which material deposited in organ systems can be transported away from the affected organs. For example, materials deposited in the respiratory tract can move out of the respiratory tract by absorption into blood, to the GI tract via the pharynx, and to regional lymph nodes via lymphatic channels (ICRP, 1994a). - pulmonary: The removal of material from the respiratory tract by particle transport and by absorption into blood (ICRP, 1994a). contamination: Radioactive material that is present in any substance, in any area, or on any surface where its presence is unwanted or unexpected. - external: Unwanted radioactive material deposited on the outside of the body on the clothing, skin, hair, body cavities such as the outer ear and eye. - internal: Unwanted radioactive material deposited within the body following an intake of the material by absorption through skin, ingestion, inhalation or through wounds. decontamination: Action taken to remove radionuclides from clothing and the external surfaces of the body, from rooms, building surfaces, equipment, or other items. decorporation: The therapeutic processes by which radioactive materials are mobilized from tissues and organs and removed from the body by enhanced material excretion. deposition: Any action resulting in the occurrence of radioactive material on or in external surfaces of the body or on or in internal tissues and organs. deterministic effects: Harmful tissue reactions that occur in all individuals who receive greater than a threshold dose; the severity of the effect varies with the dose. Examples are radiation-induced cataracts (lens of the eye), radiationinduced erythema (skin), radiation-induced pneumonitis

28 / 3. COMPENDIUM OF RADIATION FACTS AND GUIDANCE

•

• •

•

•

•

(lungs), hematopoietic failure (bone marrow), hypothyroidism (thyroid), and GI failure (GI tract). dose: A general term denoting the quantity of energy from ionizing radiation absorbed in a tissue or organ from either an external source or from radionuclides in the body. When unspecified, dose refers to quantity of absorbed dose, measured in gray (1 Gy = 1 J kg –1) or rad (1 rad = 100 ergs g –1). dose coefficient: Radiation dose per unit of activity intake. effective dose: The calculated radiation dose to the entire body, accounting for the distribution of the dose among the organs and tissues of the body, the relative biological effectiveness of the different types of radiations and for the radiation sensitivities of the various organs and tissues that might be irradiated. The term effective dose, as used in this Report for internally-deposited radionuclides, always means committed effective dose calculated over a 50 y period beyond the radionuclide intake for adults and from intake to age 70 y for intakes by children. exposure: In this Report, “exposure is the act or condition of being subject to irradiation” (ICRP, 2005a) (e.g., when a person is near a radiation source) and does not imply that external or internal radionuclide contamination has occurred, only that the potential for contamination has occurred. In the context of airborne radionuclides, exposure is the product of the air concentration of radionuclides to which a person is exposed and the duration of the exposure (ICRP, 2002a). Exposure is often used in a general sense meaning to be irradiated. When used as the specifically defined radiation quantity, exposure is a measure of the ionization produced in air by x or gamma radiation. The unit of exposure is coulomb per kilogram (C kg –1). The special unit for exposure is roentgen (R), where 1 R = 2.58 × 10–4 C kg –1. ingestion: The process in which radioactive material is taken into the digestive system. Amounts ingested are equivalent to an intake, although only a fraction may be absorbed into the blood system and deposited in tissues and organs and eventually excreted in urine. The ingested activity that is not absorbed to blood is excreted in feces. inhalation: The process in which air and substances, such as radioactive materials, entrained in the air are taken into the respiratory tract through the nose or mouth. The activity of a radionuclide inhaled may differ from the activity deposited in the respiratory tract since some fraction, depending upon its physical and chemical properties and the physiological state of the individual, may be promptly exhaled.

3.1 INTRODUCTION

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• intake: The amount of radioactive material taken into the body by inhalation, absorption through the skin, ingestion or through wounds. It is distinguished from uptake, which is the amount of material that eventually enters the systemic circulation, or deposition, which is the amount of the substance that is deposited in organs and tissues. • ionizing radiation: Electromagnetic radiation (x or gamma rays) or particulate radiation (alpha particles, beta particles, electrons, positrons, protons, neutrons, and heavy charged particles) capable of producing ions by direct or secondary processes in passage through matter. • irradiation: The action of incurring radiation by a body, tissue or organ from either external or internal radiation sources. • radionuclide: Naturally-occurring or artificially-produced unstable ion that transforms to a stable or unstable atom and releases radiations in the process. • relative biological effectiveness (RBE): For a specific radiation, the ratio of absorbed dose of a reference radiation required to produce a specific level of a response in a biological system to the absorbed dose of the specific radiation required to produce an equal response. Reference radiation normally is gamma or x rays with a mean linear energy transfer of 3.5 keV μm–1 or less. RBE generally depends on dose, dose per fraction if the dose is fractionated, dose rate, and biological endpoint. When calculating RBE-weighted absorbed doses for deterministic effects in this Report, RBE values of two and seven were used for alpha-particle irradiation of the bone marrow and lungs, respectively. • retention: Describes the propensity for radioactive materials to remain at the site of deposition. Retention is frequently described by a rate function. • stochastic effects: Effects, the probability of which, rather than their severity, is assumed to be a function of dose without a threshold. For example, cancer and hereditary effects are regarded as being stochastic • translocation: The redistribution of radionuclides from the initial sites of deposition to other tissues and organs in the body. • uptake: Quantity of a radionuclide taken up by the systemic circulation (e.g., by injection into the blood, by absorption from compartments in the respiratory or GI tracts, or by absorption through the skin or through wounds in the skin) (NCRP, 1987).

30 / 3. COMPENDIUM OF RADIATION FACTS AND GUIDANCE 3.2 Basic Radiological Facts 3.2.1

Radiation Types and Recommended Personnel Protection

• alpha: alpha radiation consists of positively charged particles emitted by certain radionuclides with a substantial amount of energy, typically 5 MeV or higher. Alpha radiation of energy less than ~7 MeV will not penetrate beyond 70 μm, the nominal dead-layer of the skin assumed in radiation protection. Alpha particles of ~4 MeV can penetrate the shallow dead-layer over some parts of the body. Alphas with energies of ≥7 MeV are emitted by only a few common radionuclides such as 212Po. Alpha-emitting radionuclides may be a health risk if taken into the body through inhalation, ingestion, or through wounds. However, most alpha emitters are not readily absorbed from the GI tract. Two important exceptions are 210Po and 226Ra, for which 10 % or more of the ingested amount may be absorbed to blood. The maximum range of a 5.3 MeV alpha particle from 239Pu in tissue is ~40 μm; thus it will penetrate on average about five cells in solid lung tissue (NCRP, 1975). (Protective actions: Wear respiratory protection to minimize inhalation or ingestion, cover all open wounds.) • beta: beta radiation consists of elementary particles emitted from nuclei during radioactive decay and have a single electrical charge. Beta radiation will penetrate protective layers of skin if the maximum beta energies are above ~70 keV, which encompasses the majority of beta-emitting radionuclides. Beta-emitting radionuclides may be a health risk if taken into the body through inhalation, ingestion, or through wounds. [Protective actions: Wear heavy clothing or turnout gear to keep high-activity and physically small beta-emitting particles (hot particles) off of skin; wear respiratory protection to minimize inhalation or ingestion.] • gamma: gamma rays are short-wave-length electromagnetic radiation. External sources of gamma rays may irradiate the entire body. High doses delivered at a high dose rate can cause an acute radiation syndrome. Gamma-emitting radionuclides taken into the body can irradiate surrounding tissues as well as those in which they are deposited. (Protective actions: Wear gloves and anti-contamination clothing to reduce skin contamination. Wear respiratory protection to minimize inhalation and ingestion. Time, distance and shielding principles can be applied to reduce external exposures from gamma-emitting radionuclides in the surrounding area.)

3.2 BASIC RADIOLOGICAL FACTS

3.2.2

/ 31

Identifying Radiation Types Using a Pancake or Other Thin End-Window Geiger-Mueller Probe Survey Meter

• A Geiger-Mueller (GM) detector with a thin window will detect alpha, beta and gamma radiation. • Each of these types of radiation has different properties which must be understood to interpret meter readings. • Background radiation (from cosmic sources or naturallyoccurring radionuclides in soil or building materials) is always present and will produce counts (and a count rate) even in the absence of radioactive contamination. • First turn on the meter in an uncontaminated area. • Subtract the background reading from all subsequent instrument readings.1 • The properties of these types of radiation can be used to help differentiate between them and to determine their relative importance. - alpha radiation will not penetrate paper or thin plastic; to screen for alpha-emitting contamination, cover the probe with paper or a plastic bag; the percent reduction in count rate (if any) is proportional to the level of alphaemitting contamination. - beta radiation will not penetrate a hand or heavy gloves; to screen for beta-emitting contamination, cover probe with hand or put inside an empty work glove; the percent reduction in count rate (if any) is proportional to the level of beta-emitting contamination. - gamma radiation will penetrate all these materials; the count rate penetrating a hand or heavy glove is proportional to the level of gamma-emitting contamination. Example: If the count rate beneath a probe is 10,000 counts per minute (cpm), the count rate through paper is 5,000 cpm, and the count rate through a glove (probe inside the glove) is 2,000 cpm; then it can be concluded that alpha-, beta- and gamma-emitting contamination is present. In this case, the gamma count rate is 2,000 cpm (the amount that penetrated both paper and glove), the beta count rate is 3,000 cpm (the amount that penetrated the paper, but not the glove), and the alpha count rate is 5,000 cpm (the amount that did not penetrate the paper or the glove). 1For example, if background radiation levels are 50 cpm and the count rate in a contaminated area is 150 cpm, the net count rate (attributed to the radiation source measured) is 150 – 50 = 100 cpm.

32 / 3. COMPENDIUM OF RADIATION FACTS AND GUIDANCE 3.2.3

Radiation Energy and Radioactive Decay Facts

The energies required to penetrate protective layer of skin (lower-energy particles will not expose living cells to radiation if the exposure is only from external sources) are: • alpha: ~7 MeV • beta: ~70 keV Average beta and positron energies: • average beta energy is about one-third the maximum energy • average positron energy is about one-half of the maximum energy Beta particle range: • in air: 3.7 m (12 feet) per million electron volts (MeV) of energy • in matter: range (g cm–2) ≈ 0.5 Emax (MeV) Range in matter depends upon the density of the matter as well as the energy of the beta particles. Range in density thickness (g cm–2) can be converted to a linear range by dividing it by the mass density of the material. For example, water has a mass density of 1 g cm–3 so a 2 MeV beta particle will travel a distance of ~1 cm in water. • range (g cm–2) ≈ 0.5 × 2 MeV = 1 g cm–2 • range (cm) ≈ range (g cm–2 ) divided by density (g cm–3) ≈ (1 g cm–2)/(1 g cm–3) = 1 cm Gamma-radiation dose rate (for point sources): · • D = 6 AEn · where D is the dose rate in rad h–1 at a distance of 1 foot, A is the source activity in curies, E is the gamma energy in million electron volts, and n is the number of gammas emitted during each disintegration (e.g., every decay of a 60Co atom emits two gammas). In a similar manner: · • D = 0.53 A E n · when D is the dose rate in rad h–1 at 1 m and A the source activity in curies.

3.3 INCIDENT RESPONSE (SECTION 18)

/ 33

Also: · • D = 0.14 A E n · when D is the dose rate in mSv h–1 at 1 m and A is the source activity in gigabecquerels. Radioactive decay: • after seven half-lives, <1 % of the original activity remains • after 10 half-lives, <0.1 % of the original activity remains Unit conversion factors: SI units and previous system: • • • • • • • • • •

1 Bq = 2.7 × 10–11 Ci = 27 pCi 1 Ci = 3.7 × 1010 Bq = 37 GBq 1 Sv = 100 rem 1 rem = 0.01 Sv 1 Gy = 100 rad 1 rad = 0.01 Gy 1 Sv Bq–1 = 3.7 × 106 rem μCi–1 1 rem μCi–1 = 2.7 × 10–7 Sv Bq–1 1 Gy Bq–1 = 3.7 × 106 rad μCi–1 1 rad μCi–1 = 2.7 × 10–7 Gy Bq–1

3.3 Incident Response (Section 18) 3.3.1

Incidents

3.3.1.1 Small Scale. Incidents occurring in laboratories, hospitals, nuclear power plants, etc., involving small amounts of radionuclides with the potential contamination of one or a few individuals are small scale. Radiation-safety and medical professionals are likely to be on the premises or nearby as employees or contract personnel. In nearly all cases, initial responding individuals will be coworkers and supervisors who are responsible for notifying radiation-safety staff/ officers (health physicists) and upper management. The radiological control staff is responsible for initial assessment and, in consultation with management, for determining the course of action for management of the contaminated person or persons, and the steps required to confine the contaminating radionuclides to the location of the incident. Appropriate local, state, or federal regulators may require notification, depending on the nature of the incident (e.g., the theft of radioactive material or the innocent finding of radioactive material).

34 / 3. COMPENDIUM OF RADIATION FACTS AND GUIDANCE 3.3.1.2 Large Scale. Incidents involving relatively-large quantities of radionuclides and the potential contamination of large numbers of people are large scale. Examples of large-scale incidents include terrorist attacks with radiological weapons, nuclear-weapons detonations, and nuclear power plant accidents. Initial responders may be law enforcement, firefighters, and local disaster response teams. Incidents may require designation of several areas, based on levels of contamination and the needs for successfully mediating the incident. The latter includes ensuring the confinement of the radionuclides to the contaminated area while effectively managing contaminated people. Some of the personnel noted below may be involved in a smallscale incident, and even large-scale incidents may not call for all of these personnel (or they may not all be available). The person(s) in charge of the incident response will be responsible for determining which roles are required and to make the best use of the available personnel and resources. 3.3.2

Roles and Responsibilities (Section 18)

Emergency medical responders: • attend to medical needs of contaminated and/or injured persons; • take actions needed to stabilize the most badly-injured individuals; • perform triage as appropriate; • transfer contaminated and injured persons for medical treatment; • exercise appropriate radiological controls; • determine which critically-injured contaminated persons should be permitted to exit without undergoing decontamination; • help decontaminate less injured persons; • perform first aid; and • help stabilize individuals requiring such attention. Radiation-safety (health-physics) responders: • establish radiation control and secured areas (Sections 4.3.3 and 18); • provide radiological coverage for all responding personnel; • issue dosimeters and record relevant information;

3.4 GUIDANCE FOR PROFESSIONALS AT INCIDENT SITE

/ 35

• control the radionuclide decontamination areas; man all control points; • conduct personal and environmental radiation contamination surveys; • supervise the use of protective clothing; track exposure of workers; • assign radiological stay-times; • supervise contamination control and decontamination efforts; • take air samples; • determine the likelihood of radioactive-material intakes; • estimate potentially-contaminated persons’ radionuclide intakes and radiation doses; and • provide advice and guidance to all nonradiation-safety personnel, including medical professionals, law enforcement, firefighters, public health, and all other responders at the scene. 3.4 Guidance for Professionals at Incident Site • Entry into a contaminated area should be made deliberately and cautiously. Determine radiation levels prior to entry, either in activity count/disintegration rate (counts per minute/disintegrations per minute) or dose rate [mGy h–1 (mR h–1)]. Wear appropriate protective clothing. • Do not attempt to enter areas in which radiation levels are >0.1 Gy h–1 (10 rad h–1) without radiation dosimeters and radiation detector OR an accompanying radiation-safety specialist, AND the concurrence of the incident commander or a designated health and safety supervisor. Radiation levels should be determined by survey with an ion chamber, micro-R meter, or energy-compensated GM detector. • Consider critical medical concerns first, followed by serious radiological concerns. • Contaminated or irradiated patients, if clothes are removed, do not pose a health risk to emergency responders unless the patient contains embedded fragments of high-activity radioactive material. Health-care workers have not been injured by a contaminated patient. • Take precautions to minimize the risk of spreading contamination.

Radiation Readings and Their Significance (dose-rate meters) TABLE 3.1—Interpretation of dose-rate meter readings. Dose

20 μGy h

–1

Ratea

Significance –1

(2 mR h )

• Regulatory limit for radiation dose rate in an uncontrolled area

• Evacuate public if practicable • Establish control boundary

50 μGy h–1 (5 mR h–1)

• Regulations require posting as radiation area

• Entry requires wearing dosimetry

1 mGy h–1 (100 mR h –1)

• Regulations require posting as high radiation area

• Entry requires double dosimetry • Nonemergency turn-back limit

• Radiation workers will exceed annual dose limit [50 mSv (5 rem)] in 30 min

• Turn-back limit except for grave emergency or lifesaving

• May develop radiation sickness with 1 h exposure

• Turn-back limit except for lifesaving, minimize time in these areas

100 mGy h–1 (10 R h–1) 1 Gy h–1 (100 R h–1) 10 Gy h–1 (1,000 R h–1)

a

Actions

• Do not enter except under the most severe • May develop radiation sickness with 5 – circumstances 6 min exposure • Will reach LD50 dose (the lethal dose to 50 % of the population) in 30 min • Will receive lethal dose in 1 h

NCRP Report No. 138 (NCRP, 2001a) recommends a maximum radiation dose of 500 mSv (50 rem) for those engaged in emergency actions during a radiological emergency.

36 / 3. COMPENDIUM OF RADIATION FACTS AND GUIDANCE

3.4.1

3.4.2

Surface Radiation Readings and Their Significance (contamination survey meters) TABLE 3.2—Interpretation of contamination survey readings (adapted from DOE, 1999). Significance

Actions

0.33 Bq (20 dpm); alpha

Lower limit for alpha emitters

• No control

0.33 Bq (20 dpm); alpha or 16 Bq (1,000 dpm); beta/gamma

Limit for contamination in uncontrolled area

• Establish control boundary • Wear anti-contamination clothing

33 Bq (2,000 dpm); alpha or 1,670 Bq (100,000 dpm); beta/gamma

Resuspended contamination may lead to inhalation or ingestion risk

• High control • Wear anti-contamination clothing • Consider respiratory protection

16.7 kBq (1,000,000 dpm); alpha, beta, gamma

Likely to lead to inhalation of significant amounts of activity

• Very-high control procedures required • Wear anti-contamination clothing and respiratory protection

a Instruments that readout in counts per minute or counts per second require calibration to give becquerel or disintegrations per minute. For example: dpm = counts per minute divided by counting efficiency (in counts per disintegration) Bq = counts per second divided by counting efficiency (1 Bq = 60 dpm)

3.4 GUIDANCE FOR PROFESSIONALS AT INCIDENT SITE

Contamination (Bq or dpm 100 cm–2)a

/ 37

38 / 3. COMPENDIUM OF RADIATION FACTS AND GUIDANCE 3.5 Management of Potentially-Injured and Contaminated Persons 3.5.1 1. 2.

3. 4. 5. 6. 7.

8. 9. 10. 3.5.2

Priorities for Aiding Contaminated Individuals life-threatening injuries and medical conditions; high levels of internal contamination (thousands of disintegrations per minute in nasal or oral swabs from ingested or inhaled activity, or thousands of disintegrations per minute from contaminated wounds); evidence of radioiodine or transuranic intakes; serious injuries (not life threatening, but requiring rapid medical attention); high levels of skin contamination (hundreds of thousands of disintegrations per minute or more); moderate injuries (requiring medical attention); moderate levels of internal contamination (hundreds of disintegrations per minute in nasal or oral swabs or from contaminated wounds); moderate levels of skin contamination (thousands to hundreds of thousands of disintegrations per minute); mild injuries with low levels of contamination [<1,000 disintegrations per minute (dpm)]; and contaminated but uninjured persons. Stages in Management of Exposed Persons

REACT/TS (ORAU, 2009) and the Office of the Assistant Secretary for Preparedness and Response, National Library of Medicine, DHHS (2009) have prepared decision-tree charts for medical responses to radiological and nuclear incidents. Drawing largely from these sources, a decision tree (Figure 3.1) is provided to guide medical and radiation safety professionals through nine stages in the management of persons exposed to radionuclides. The first three of nine stages can be accomplished onsite. The subsequent four stages should occur remote from the contaminated area, preferably at a hospital, clinic, or other medical emergency facility. The final two stages will occur post-hospital at a site depending upon the circumstances. The nine stages in managing contaminated persons are not necessarily sequential. Depending upon the nature of the contamination and possible injuries, certain stages may be bypassed. These stages suggest the proper course of action to be taken to achieve a favorable outcome for exposed and contaminated individuals. It is recommended that the activities performed and the information obtained at each stage be documented and that

3.5 MANAGEMENT OF CONTAMINATED PERSONS

/ 39

Fig. 3.1. Decision tree for management of persons contaminated with radionuclides (adapted from DHHS, 2009; ORAU, 2009).

40 / 3. COMPENDIUM OF RADIATION FACTS AND GUIDANCE copies accompany the patient. These nine stages illustrated in Figure 3.1 are color coded to appropriate parts and sections of the Handbook for specific guidance. Stage 1. Medical Assessment (Section 6): Onsite Triage Area Screen potentially-exposed persons for life-threatening problems and for evidence of exposure and contamination. Depending upon the nature of the contaminating incident, it may be that not all persons at the site will be exposed and contaminated. This assessment should occur at the site of the incident. • Examine exposed persons for signs of injuries and potential life-threatening problems. This has priority over assessing contamination and decontamination. • Stabilize life-threatening problems. • Observe for signs of psychological distress because psychosocial effects can be both widespread and long lasting constituting some of the most significant and challenging consequences of a radiation incident (Becker, 2001) (Sections 9.1.2 and 13.4). • Examine exposed persons for signs of acute radiation exposure. Table 6.2 describes early effects of whole-body irradiation from external and internal sources. • Survey the individuals’ facial areas and check nasal swabs with a GM, sodium iodide (gamma radiation), or zinc-sulfide (alpha radiation) detector for evidence of inhalation or ingestion. Consider prompt treatment with KI if high intakes of radioiodine are suspected and with DTPA if intakes of transuranics are indicated (Section 12). • Survey wounds with a GM, sodium iodide (gamma radiation), or zinc-sulfide (alpha radiation) detector for evidence of contamination or shrapnel and flush with sterile saline solution or water if contamination is detected or suspected (Section 9.4.1). • Attempt to determine the time of intake as accurately as possible. • Air samples and any other samples such as swabs of contaminated surfaces should be collected for analysis. Stage 2. External Contamination Assessment (Section 7): Onsite Triage Area In a controlled area, perform whole-body surveys and record the levels and locations of the contaminating radionuclides.

3.5 MANAGEMENT OF CONTAMINATED PERSONS

• • • • •

/ 41

locate contaminated body area, including orifices; identify hot particles, shrapnel and contaminated debris; identify and quantify radionuclides; evaluate potential for skin injury; and confirm internally-contaminated individuals.

Stage 3. External Decontamination (Section 8): Onsite Decontamination Area In a controlled area, remove clothing and decontaminate all areas showing evidence of contamination until levels are less than twice the background reading (Section 8.1.1) and reduce amounts of radionuclides in wounds. Stage 4. Patient Evaluation and Emergency Care (Section 9): Hospital Evaluation of injuries and emergency care of contaminated persons in hospital; guidance in controlling contamination: • treat patients with injuries; • evaluate patients for evidence of radiation sickness and begin treatment; • evaluate patients with skin burns, contaminated wounds, and radionuclide intakes for emergency treatment; • consider treatment with KI if intakes of radioiodine are suspected and DTPA if intakes of transuranics are indicated; and • observe patients for adverse psychosocial reactions. Stage 5. Internal Contamination Assessment (Section 10): Hospital Collection and evaluation of bioassay data such as in vivo counting, nasal swabs, and excreta should occur as soon as possible in an emergency facility or hospital. Estimate radionuclide intakes and radiation doses. Stage 6. Clinical Decision Guidance (Section 11): Hospital Evaluate internal contamination and radiation dose with respect to the CDG. Stage 7. Medical Management (Section 12): Hospital This will generally occur in hospital if, after evaluation of the intake, the physician recommends treatment.

42 / 3. COMPENDIUM OF RADIATION FACTS AND GUIDANCE

Stage 8. Follow-Up Medical Care (Section 13): Location Depends Upon Circumstances Patients should be followed for evidence of latent development of radiation injuries, monitoring internal contamination, possible epidemiology studies, and negative psychosocial reactions. Stage 9. Contaminated Decedents (Section 14): Hospital and Mortuary Guidance is provided to protect medical examiners and mortuary professionals from radiation exposure and to control radionuclide contamination of individuals and facilities. 3.6 Radiation Exposures from External Sources 3.6.1

Health Effects from External Radiation

Table 3.3, adapted from ICRP Publication 96 (ICRP, 2005a), may be useful in assessing the potential health risks of such wholebody doses. This table provides a summary of the health effects that can result from whole-body x- or gamma-radiation doses ranging from background to very-high dose levels. 3.6.2

Neutron-Radiation Dose from Criticality Accident (based on 24Na activation in body)

Naturally-occurring 23Na is activated by neutron exposure to produce 24Na, which then emits 1.37 and 2.87 MeV gamma rays. The half-life of 24Na is 15 h. Estimating neutron dose using the 24Na “quick sort” method (Tables 3.4 and 3.5): • Place probe of radiation monitoring instrument against the abdomen with the person bent over during the measurement (the armpit may be used if person is unable to bend over or external contamination prohibits such a measurement). • If the detector (e.g., Cutie-pie type ionization chamber) is used to measure radiation dose rate, the following relationships can be used to determine the neutron dose: · 3,600 ( D mGy ) D Gy = ------------------------------------------body weight (kg) · 8,000 ( D mrad ) D rad = -----------------------------------------body weight (lb)

(3.1)

TABLE 3.3—Simplified summary of radiation-induced health effects from whole-body radiation (adapted from ICRP, 2005a). Expected Dose

Effects

Outcome

No acute effect, extremely small additional cancer risk (<0.1 %).

No observable increase in the incidence of cancer, even in large exposed population (~1,000,000 people).

Low dose: towards 100 mSv (10 rem) effective dose.

No acute effects, subsequent additional cancer risk of <1 %.

Possible observable increase in the incidence of cancer, if the exposed group is large (perhaps greater than ~100,000 people).a

Moderate dose: towards 1,000 mSv (100 rem) (acute whole-body dose).

Nausea, vomiting possible, mild bonemarrow depression; subsequent additional cancer risk of ~10 %.

Probable observable increase in the incidence of cancer, if the exposed group is more than a few hundred people.

High dose: >1,000 mSv (100 rem) (acute whole-body dose).

Certain nausea, likely bone-marrow syndrome, high risk of death from ~4,000 mSv (400 rem) of acute whole-body dose without medical treatment. Significant additional cancer risk.

Observable increase in the incidence of cancer.

aEpidemiological

studies do not have the power to reveal cancer risks in the dose range up to ~100 mSv (10 rem) (ICRP, 2007).

3.6 RADIATION EXPOSURES FROM EXTERNAL SOURCES

Very-low dose: ~10 mSv (1 rem) effective dose or less.

/ 43

44 / 3. COMPENDIUM OF RADIATION FACTS AND GUIDANCE TABLE 3.4—Survey instrument readings on the body after an criticality accident (SI units) (Rathbone, 2007).

Weight of Subject (kg)

Pancake GM net Reading (cpm) for a 0.1 Gy Neutron Dose

Dose-Rate Instrument Reading (μGy h–1) for a 1 Gy Neutron Dosea

Time Following Exposure (h) 0

4

15

0

4

15

60

600

500

300

17

4

15

70

700

580

350

19

16

10

80

800

660

400

22

18

11

90

900

750

450

25

21

12

100

1,000

830

500

28

23

14

110

1,100

910

550

31

25

15

aAir-filled

ion chamber.

TABLE 3.5—Survey instrument readings on the body after an criticality accident (previous system) (Rathbone, 2007).

Weight of Subject (lb)

Pancake GM net Reading (cpm) for a 10 rad Neutron Dose

Dose-Rate Instrument Reading (mrad h–1) for a 100 rad Neutron Dosea

Time Following Exposure (h) 0

4

15

0

4

15

125

570

470

280

1.6

1.3

0.8

150

680

570

340

1.9

1.6

0.9

175

800

660

400

2.2

1.8

1.1

200

910

760

450

2.5

2.1

1.2

225

1,000

850

510

2.8

2.3

1.4

250

1,100

940

570

3.1

2.6

1.6

a

Air-filled ion chamber.

3.6 RADIATION EXPOSURES FROM EXTERNAL SOURCES

/ 45

· where D is the radiation dose-rate reading in milligray (millirad) per hour, using a probe calibrated with 137Cs. • If the detector (e.g., GM pancake-type radiation monitoring instruments) is used to measure radiation count rate (cpm), the following relationships can be used to determine the neutron dose: 0.01 (cpm) D Gy = -------------------------------------------body weight (kg)

(3.2)

2.2 (cpm) D rad = ------------------------------------------body weight (lb) Notes: • Decay correction is not necessary if the measurement is made within the first few hours after exposure (IAEA, 1974; Rathbone, 2007). • The radiation dose calculated by this method only addresses the neutron component. Gamma-to-neutron ratios or personal gamma dosimetry must be used to determine the dose from gamma radiation. Example 1: A 75 kg person was exposed to neutron radiation and measures 750 cpm above background using a GM pancake-type detector. The calculated radiation dose is 0.1 Gy (10 rad). Example 2: A 200 lb person was exposed to neutron radiation. Using the quick-sort method, a Cutie-pie-type ionization chamber is used to measure radiation dose rate, which is 0.022 mGy h–1 (2.2 mrad h–1). The calculated radiation dose is 0.88 Gy (88 rad). 3.6.3

Exposures from Sealed Radioactive Sources

In many situations it may be possible to determine the activity of a radioactive source by referring to shipping documents, source documentation, asking the source’s owner, contacting the source manufacturer, or by other means such as measurements. If the source activity can be ascertained, the information in Table 3.6 can be used to help determine the risk that the source may pose to bystanders and response personnel. In some cases, it may not be possible to determine the activity of a radioactive source but potential whole-body radiation doses to

46 / 3. COMPENDIUM OF RADIATION FACTS AND GUIDANCE TABLE 3.6—Activity and significance of sealed radioactive sources. Activity

Significance

Actions

10s of kBq (μCi)

Not a significant health risk

• Control as radioactive material

10s of MBq (mCi)

May be a minor health risk

• Control as radioactive material, do not handle directly, use tongs

10s of GBq (Ci)

May be a health risk

• Do not handle directly, use tongs

100s of GBq (a few Ci)

May be a health risk

• Do not approach closely • Do not handle source • Establish 20 μGy h–1 (2 mrad h–1) boundary

TBq (10s of Ci)

May be a severe health risk

• Do not approach without radiation dose and dose-rate monitoring equipment • Establish 20 μGy h–1 (2 mrad h–1) boundary

bystanders and response personnel may be estimated from measurements of the air-kerma rate (exposure rate) from the source and comparing them with the values in Table 3.7. 3.7 Air Kerma and Skin Doses for Point Sources 3.7.1

Intervention Levels for Skin Contamination

Deterministic effects are estimated to occur at skin doses of 2 to 3 Gy (200 to 300 rad), such as from beta emitters deposited on the skin. With respect to “hot particles” on the skin, NCRP concluded that the risk of stochastic effects was negligible compared with deterministic effects. It was recommended that the dose to skin (and the ear) at depth of 70 μm be limited to 0.5 Gy (50 rad) averaged over the most highly exposed 10 cm2 area of skin. This is a limit per particle, with no overlap. For the eye and respiratory tract (anterior nose) the limits are annual because of the small mass of tissue. For the eye, the recommended limit for hot particles is 5 Gy (500 rad) at 70 μm averaged over the most highly exposed 1 cm2 of ocular tissue. For hot particles sequestered in the anterior nasal

TABLE 3.7—Air-kerma rate (1 m from point source) and electron constants (skin dose rate at depth of 70 μm from source on skin surface) for selected radionuclides. Half–Life

Decay Modec

Air-Kerma Rated [Gy s–1 (Bq m–2)–1]

60

Co

5.27 y

β–

8.53 × 10–17

90

Sr

28.8 y

β–

*90Y

64.1 h

β–

1.65 × 10–21

—

Electron Constante

[rad h–1 (μCi m–2)–1]

[Gy s–1 (Bq cm–2)–1]

[rad h–1 (μCi cm–2)–1]

1.14 × 10–6

3.09 × 10–10

4.12 × 100

—

4.98 × 10–10

6.63 × 100

2.20 × 10–11

6.65 × 10–10

8.86 × 100

I

8.0 d

β–

1.45 × 10–17

1.93 × 10–7

4.82 × 10–10

6.42 × 100

137

Cs

30.2 y

β–

6.11 × 10–23

8.14 × 10–13

4.73 × 10–10

6.30 × 100

*137mBa

2.55 m

IT

2.26 × 10–17

3.01 × 10–7

6.88 × 10–11

9.16 × 10–1

192

Ir

73.8 d

β– EC

3.18 × 10–17

4.24 × 10–7

5.46 × 10–10

7.27 × 100

210

Po

138. d

α

3.60 × 10–22

4.80 × 10–12

8.53 × 10–17

1.14 × 10–6

235U

7.04 × 108 y

α

1.32 × 10–17

1.76 × 10–7

4.69 × 10–11

6.25 × 10–1

238

U

4.47 × 109 y

α SF

2.04 × 10–18

2.72 × 10–8

2.70 × 10–13

3.60 × 10–3

*226Ra

1,600 y

α

5.23 × 10–19

6.97 × 10–9

1.42 × 10–11

1.89 × 10–1

87.7 y

α SF

2.64 × 10–18

3.52 × 10–8

2.49 × 10–13

3.32 × 10–3

2.41 × 104 y

α

1.11 × 10–18

1.48 × 10–8

1.29 × 10–13

1.72 × 10–3

238Pu 239

Pu

/ 47

131

3.7 AIR KERMA AND SKIN DOSES FOR POINT SOURCES

Radionuclideb

Radionuclideb

Half–Life

Decay Modec

Air-Kerma Rated [Gy s–1 (Bq m–2)–1]

[rad h–1 (μCi m–2)–1]

Electron Constante [Gy s–1 (Bq cm–2)–1]

[rad h–1 (μCi cm–2)–1]

*241Am

432 y

α

9.80 × 10–18

1.31 × 10–7

7.60 × 10–13

1.01 × 10–2

252

2.65 y

α SF

7.54 × 10–17

1.00 × 10–6

1.32 × 10–10

1.76 × 100

Cf

a

See Section 7 for additional radionuclides and Section 21.3 for explanation of skin dose calculations. preceded by an asterisk are radioactive progeny that may be present in significant quantities. c EC = electron capture IT = isomeric transition SF = spontaneous fission dThe air-kerma rates include the contribution of annihilation radiations associated with the emission of positrons and neutrons, prompt gamma, and delayed gamma associated with spontaneous fission. eThe electron constant includes the contribution from delayed beta emissions associated with spontaneous fission. bNames

48 / 3. COMPENDIUM OF RADIATION FACTS AND GUIDANCE

TABLE 3.7—(continued)

3.8 RADIATION EXPOSURES FROM INTERNAL DEPOSITIONS

/ 49

passages, the recommended limit is 5 Gy (500 rad) at 70 μm averaged over the most highly exposed 1 cm2. With the possible exception of situations where hot particles are continually moving within tissues and irradiating an increasing volume of tissues and numbers of cells, hot particles are a greater risk of causing deterministic effects than stochastic effects such as cancer (NCRP, 1999). Default operational intervention levels for skin contamination suggested by the International Atomic Energy Agency (IAEA, 2005a) are given in Table 3.8. Appropriate actions for each level are indicated in the fourth column (Section 7). 3.7.2

Guidance for Decontaminating Skin

The skin decontamination objective is to reduce the level to less than two times background by washing the skin. The number of washings should be limited to avoid skin injury; two cycles or as long as each washing reduces the level by 50 % (Section 8). When the lack of facilities or equipment or the number of contaminated people make it impracticable to achieve this objective, other guidance may apply as shown in Table 3.9. 3.8 Radiation Exposures from Internal Depositions of Radionuclides 3.8.1

Health Effects from Internal Radionuclide Contamination

3.8.1.1 Deterministic Effects (harmful tissue reactions). Table 3.10 summarizes the organs and tissues at risk for deterministic effects if sufficient quantities are taken into the body through the respiratory tract. If these radionuclides are ingested in sufficient quantities (especially the beta/gamma emitters) the GI tract is at risk as well as the other tissues in the body in which the radionuclides may deposit. If taken into the body through wounds in the skin, the tissues at risk from the more soluble radionuclides are those in which the radionuclides tend to deposit, as shown in the table. Ingested insoluble alpha emitters are a minimal risk to the body because little of the alpha radiation penetrates to the sensitive cells in the GI tract. Insoluble material entering wounds tends to be sequestered in lymphatic tissues. In these cases the radiation dose from alpha emitters is localized, while that from beta/gamma emitters may penetrate further into adjacent tissues. Deterministic effects observed following human intakes of radionuclides are very limited, primarily because the large intakes required to cause deterministic effects have been rare. These include thyroid nodules in Marshall Islanders exposed to radioiodine in the fallout from nuclear-weapon tests in the Pacific (Mettler and Upton,

Alpha {Bq cm–2 (nCi cm–2) [dpm cm–2]}

Beta/Gamma {Bq cm–2 (nCi cm–2) [dpm cm–2]}

Beta/Gamma (low background area)a [μSv h–1 (μrem h–1)]

<10 (<0.27) [<600]

<100 (<2.7) [<6,000]

Not detectable

None • allow release

>10 (>0.27) [>600]

>100 (>2.7) [6,000]

Not detectable

Intervention optional • decontaminate or advise to shower and wash clothing • no significant health risk • slow release

>100 (>2.7) [>6,000]

>1,000 (>27) [>60,000]

0.2 – 0.3 (20 – 30)

Intervention advisable • prevent inadvertent ingestion and inhalation, limit spread of contamination and decontaminate

>1,000 (>27) [>60,000]

>10,000 (>270) [>600,000]

2–3 (200 – 300)

Intervention required • prevent inadvertent ingestion and inhalation, limit spread of contamination and decontaminate

aAmbient

dose equivalent rate measured at 10 cm from skin surface.

Actions

50 / 3. COMPENDIUM OF RADIATION FACTS AND GUIDANCE

TABLE 3.8—Skin contamination intervention levels (adapted from IAEA, 2005b).

3.8 RADIATION EXPOSURES FROM INTERNAL DEPOSITIONS

/ 51

TABLE 3.9—Decontamination guidance; applicable when large numbers of people are contaminated and the goal of less than two times background is impractical. Spota (0.2 cm2)

Body Surface

Alphab

<0.37 kBq (0.01 μCi) 22,000 dpm

17 Bq cm–2 (0.45 nCi cm–2) 1,000 dpm cm–2

IAEA (2005b)

Beta/gamma

<3.7 kBq (0.1 μCi) 220,000 dpm

<170 Bq cm–2 (4.5 nCi cm–2) 10,000 dpm cm–2

FEMA (2002) NCRP (2005a)

Contamination

Reference

aFor contamination fixed on skin. Limit is factor of 10 greater for mixed loose and fixed. bThe value for alpha radiation is a factor of 10 less than beta/gamma radiations consistent with IAEA skin contamination operational intervention levels.

TABLE 3.10—Summary of organs or tissues at risk from deterministic effects following inhalation of example radionuclides (Table 16.13). Radionuclide and Form

Emissions

Absorption Typea

Organs/Tissues at Riskb

β

F (fast)

Bone marrow

β

F (fast)

Thyroid

137CsCl

βγ

F (fast)

Bone marrow

144CeO 2

βγ

S (slow)

Lungs

90

SrCl2

131

I vapor

210

Po

α

F or M (fast or moderate)

Lungs, bone marrow, kidneys, liver, others

238

PuO2

α

M (moderate)

Lungs, thoracic lymph nodes, liver, bone

α

S (slow)

Lungs, thoracic lymph nodes

239PuO

aBased

2

on rate of absorption into blood from respiratory tract (ICRP, 1994a). Organs and tissues at risk are those in which the radionuclides are preferentially deposited and retained, in some cases, for relatively long periods. b

52 / 3. COMPENDIUM OF RADIATION FACTS AND GUIDANCE 1995; NCRP, 2008). Early death occurred in a person following the ingestion of a large quantity of 210Po (Harrison et al., 2007). The ingestion of large quantities of 137Cs resulted in the deaths of four persons and serious injuries in a number of others in Goiânia, Brazil (IAEA, 1988; 1998a). Deaths from radiation pneumonitis/pulmonary fibrosis occurred in early plutonium workers at the Mayak Production Association in the former Soviet Union (Claycamp et al., 2000). 3.8.1.2 Stochastic Effects 3.8.1.2.1 Cancer. Tissues at risk for the development of cancer later in life are primarily those in which radionuclides are deposited and retained for extended periods. Examples are the tissues listed in Table 3.11 for several radionuclides taken into the body through the respiratory tract. The same tissues are at risk when the route of entry is through the GI tract or through the skin except for those insoluble forms that are retained in the respiratory tract. Epidemiology studies have identified statistically significant increases in cancer incidence in only a few populations of the many that have been exposed to natural and man-made radionuclides through the years and extensively studied. Studies of workers at former Soviet Union nuclear weapons facilities and nearby residents are still in progress (UNSCEAR, 2008). Those showing increases include: • lung cancer in underground hard rock and uranium miners exposed to radon; • bone cancer in radium dial painters; • bone and liver cancers in patients treated with 232Th and progeny [Thorotrast® (VanHeyden Company, DresdenRadebeul, Germany)], lung, bone and liver cancers in plutonium workers at the Mayak Production Association; • thyroid cancers in individuals exposed as children at the Chernobyl nuclear reactor accident; • thyroid cancer in patients treated with 131I for thyroid disorders; and • bone cancers in ankylosing spondylitis patients treated with 224Ra (UNSCEAR, 2000; 2008) (Section 16.7.2.1). 3.8.1.2.2 Hereditary effects (Section 16.7.2.2). Hereditary effects have not been considered a serious risk following intakes of radionuclides. 3.8.1.3 Developmental Effects (Section 16.7.3). The risk of developmental effects occurring as a result of internal contamination is

3.8 RADIATION EXPOSURES FROM INTERNAL DEPOSITIONS

/ 53

TABLE 3.11—Tissues at risk for cancer induction by radionuclides taken into the body through the respiratory tract in a chemical form with the clearance characteristics specified (Sections 17.7.2 and 20 provide references and further information).a Radionuclide 3H

2O

Emissions

Tissues at Riskc

Absorption Typeb

β

F (fast)

Total body

60

Co oxides

γ

S (slow)

Lung

90

SrCl2

β

M (moderate)

Bone marrow, bone

β

S (slow)

Lung

β

F (fast)

Thyroid

137CsCl

γ

F (fast)

Total body

144CeO 2

β

S (slow)

Lung

210

α

M (moderate)

Lungs, spleen, kidney, blood cells, liver, bone marrow

Rn and decay products

α

F (fast)

Lung

226Ra

α

M (moderate)

Bone

α

S (slow)

Lung

106RuO 131

2

I vapor

Po

222

232

sulfate

Th

238PuO

2

α

M (moderate)

Lung, bone, liver

239PuO

2

α

S (slow)

Lung,

α

M (moderate)

Bone, lung

241

AmO2

a Natural and depleted uranium are not included because UNSCEAR concludes “… there is little or no epidemiological evidence for an association between uranium and any cancer.” (Volume I, Annex A of UNSCEAR, 2008). b Based on rate of absorption into blood from the respiratory tract (ICRP, 1994a). cOrgans and tissues at risk are those in which the radionuclide preferentially deposits and is retained.

54 / 3. COMPENDIUM OF RADIATION FACTS AND GUIDANCE considered minimal because the exposure of the fetus to at least 0.1 Gy (10 rad) would have to occur three to eight weeks after conception. 3.8.2

Inhalation Intakes

Only in very extreme situations can the inhaled amounts approach levels that would cause early health effects (deterministic effects). Relatively-high air concentrations of radionuclides would have to be sustained for many minutes or even hours for such intakes to occur (Section 16). 3.8.2.1 Air Samples. Air samples can be useful as indicators of exposures to airborne radionuclides and in estimating inhalation intakes. Personal air samplers are used in many occupational situations where airborne radionuclides are relatively frequent. These are generally accompanied by room or area air samplers. In incidents where large numbers of people may be contaminated, air samples are not likely to be collected, if at all, until sometime after the incident. When air samples are available, they can be used to assess the potential for deterministic effects. Table 3.12 gives estimates of the air concentrations of several radionuclides that would have to be breathed for 10 min to result in intakes that would give radiation doses approximating threshold levels for deterministic effects and effective doses of 0.25 Sv (25 rem) (see Section 16 for further information). 3.8.2.2 Nasal Swabs • Particle-size distribution of the inhaled material is an important factor. For aerosols with an AMAD of 5 μm, ~34 % of the inhaled amount is deposited in the anterior nasal passages of an adult working male, whereas only ~17 % is deposited when the aerosol has an AMAD of 1 μm (ICRP, 1994a). The portion of inhaled activity collected on nasal swabs in the early hours after inhalation of a radionuclide is highly variable, depending upon such factors as aerosol particle size, the extent of nose versus mouth breathing during the exposure, and the amount of nose blowing and wiping of the nostrils since the beginning of the exposure. Based on model predictions and controlled human experiments, it is estimated that the combined activity on swabs of both nostrils represents ~5 % of the amount inhaled. See Section 10.3.1.1 for guidance in taking and interpreting nasal swabs and Section 16.4.1.2 for additional information.

3.8 RADIATION EXPOSURES FROM INTERNAL DEPOSITIONS

/ 55

TABLE 3.12—Estimates of the concentrations in air [MBq m–3 (μCi m–3)] of several radionuclides that would have to be inhaled for 10 min to achieve intakes sufficient to produce deterministic effects or give effective doses of 0.25 Sv (25 rem) (Section 16.7). Radionuclidea

90 SrCl2 (Type F) 131I (Vapor) 137CsCl (Type F) 144

CeO2 (Type S) 210PoCl 2 210 PoCl4

Air Concentrations Required to Cause Deterministic Effectsb,c

Air Concentrations Required to Result in an Intake of 1 CDG Leading to an Effective Dose of 0.25 Sv (25 rem)c

2,600 (70,000) Bone-marrow depression

51

(1,400)

30 (800) Hypothyroidism

76

(2,100)

8,000 (220,000) Bone-marrow depression

350

(9,500)

3,700 (100,000) Pneumonitis

52

(1,400)

1,900 (51,000) Bone-marrow depression

0.67

(18)

PuO2 (Type M)

40 (1,100) Pneumonitis

0.049

(1.3)

239PuO

40 (1,100) Pneumonitis

0.18

(4.9)

40 (1,100) Pneumonitis

0.57

(1.5)

or

(Type M) 238

2

(Type S) 241

AmO2 (Type M) aThe

radionuclides shown here are used as examples to demonstrate the levels of airborne activity required to cause serious health concerns. Assumed a breathing rate of 1.2 m3 h–1 of unfiltered air for an adult and a lognormal particle-size distribution with AMAD = 5 μm. b Deterministic effects expressed within two to three months are given for the particular radionuclide. c Calculations based on a breathing rate of 1.2 m3 h–1 of unfiltered air by an adult, 5 μm AMAD particles, and a total deposition of 82 % (ICRP, 1994a).

• Other complicating factors include uncertainty about the length of time since inhalation, whether the individual was mouth breathing, and whether the individual has respiratory problems (e.g., sinus restrictions, upper respiratory infections, etc.). • High activity on a nasal swab does not always imply high penetration to lower regions of the respiratory tract.

56 / 3. COMPENDIUM OF RADIATION FACTS AND GUIDANCE 3.8.2.3 Doses Received from Inhaled Radionuclides. The following table, Table 3.13, is derived from data in Section 20 on absorbed and effective dose2 coefficients for radionuclides (5 μm AMAD particle-size distribution unless otherwise specified). 3.8.3

Intakes Through Skin and Ingestion

Wounds in areas of skin contamination are strong indicators of possible radionuclide intakes. More detailed information on the behavior of radionuclides in wounds and possible treatments is given in NCRP Report No. 156 (NCRP, 2006a). Contamination of skin surfaces suggest possible intakes through absorption, but only in cases of very heavy contamination would absorption result in significant internal contamination, even if the radionuclide is in a soluble form (Section 10). Injuries with contaminated debris and shrapnel are clear evidence of internal contamination. Discovery of contaminated food and water should be taken as evidence of possible ingestion intakes. Contamination of the mouth and other oral surfaces suggest possible intakes by ingestion, not necessarily in contaminated food, but by touching the face and mouth with contaminated hands. 3.9 Medical Management of Internal Radionuclide Depositions 3.9.1

Clinical Decision Guides

The Clinical Decision Guide (CDG), a new operational quantity, is defined in Section 11 to provide a basis that physicians can use when considering the need for medical treatment for internallydeposited radionuclides or as a screening level indicating the need for a more detailed investigation of tissue-specific absorbed doses over different time periods. For radionuclides other than isotopes of iodine, CDG is the maximum, once-in-a lifetime intake of a radionuclide that represents: (1) a stochastic risk, as judged by the calculated effective dose over 50 y for intake by adults and to age 70 y for intake by children that is in the range of risks associated with guidance on dose limits for emergency situations (DOE, 2008a; FEMA, 2008; ICRP, 1991a; NCRP, 1993; 2005a); and (2) avoidance of deterministic effects as judged by the calculated 30 d RBE-weighted 2The term effective dose, as used in this Report for internally-deposited radionuclides, always means committed effective dose calculated over a 50 y period beyond the radionuclide intake for adults and from intake to 70 y of age for intakes by children.

TABLE 3.13—Dose estimates for inhalation intake of radionuclides by adults (Section 20). Absorbed Dose (30 d) Radionuclidea

227Ac 241

Am

144

D

Ce, D

137Cs,

D

60Co

Method of Measurementd

Absorption Typee

21.8 y

αβγ

NS, IVC, U, F

Type M

432 y

αγ

2.65 y

αγn

285 d

βγ

Effective Dose (50 y)

Tissues (Gy Bq–1)

(rad μCi–1)

(Sv Bq–1)

(rem μCi–1)

Lung

8.2 × 10–7

3.0 × 100

4.8 × 10–5

1.8 × 102

2.3 × 10

2.7 × 10

–5

1.0 × 102

1.1 × 10–5

4.1 × 101

NS, IVC, U, F

Type M

Lung

6.3 × 10

NS, BC, U

Type M

Lung

9.7 × 10–7

3.6 × 100

Red marrow

1.3 × 10–8

4.8 × 10–2

Type M

Lung

–8

4.1 × 10

1.5 × 10–1

2.3 × 10–8

8.5 × 10–2

Type S

Lung

4.9 × 10–8

1.8 × 10–1

2.9 × 10–8

1.1 × 10–1

NS, BC, U

–7

0

30.2 y

βγ

NS, BC, U

Type F

Red marrow

1.5 × 10–9

5.6 × 10–3

4.3 × 10–9

1.6 × 10–2

5.27 y

βγ

NS, BC, U

Type M

Lung

1.5 × 10–8

5.6 × 10–2

7.1 × 10–9

2.6 × 10–2

Type S

Lung

1.7 × 10–8

6.2 × 10–2

1.7 × 10–8

6.3 × 10–2

244Cm

18.1 y

αγn

NS, IVC, U

Type M

Lung

7.3 × 10–7

2.7 × 100

1.7 × 10–5

6.3 × 101

154Eu

8.59 y

β, EC

NS, IVC, U, F

Type M

Lung

2.5 × 10–8

9.2 × 10–2

3.2 × 10–8

1.2 × 10–1

Type S

Lung

2.9 × 10–8

1.1 × 10–1

2.4 × 10–8

8.8 × 10–2

U

HTO

Red marrow

1.5 × 10–11

5.6 × 10–5

1.8 × 10–11

6.7 × 10–5

3H 131I, 192

β

8.02 d

βγ

IVC, U

Vapor

Thyroid

4.1 × 10–7

1.5 × 100

2.0 × 10–8

7.4 × 10–2

73.8 d

βγ

NS, BC, U, F

Type M

Lung

1.7 × 10–8

6.3 × 10–2

4.1 × 10–9

1.5 × 10–2

/ 57

Ir

D

12.3 y

3.9 INTERNAL RADIONUCLIDE DEPOSITIONS

252Cf,

Emissionsc

Half-Lifeb

Absorbed Dose (30 d) Radionuclidea

Half-Lifeb

Emissionsc

P

14.3 d

β

238Pu

87.7 y

αn

32

239

Pu

24,110 y

210Po

226Ra,

138 d

D

103Ru

106Ru

153

Sm

90Sr,

D

α

α

Method of Measurementd

Absorption Typee

IVC, U

Type M

IVC, U, F

IVC, U, F

U

Effective Dose (50 y)

Tissues (Gy Bq–1)

(rad μCi–1)

(Sv Bq–1)

(rem μCi–1)

Lung

1.4 × 10–8

5.2 × 10–2

2.9 × 10–9

1.1 × 10–2

Type M

Lung

6.3 × 10–7

2.3 × 100

3.1 × 10–5

1.1 × 102

Type S

Lung

7.4 × 10–7

2.7 × 100

1.1 × 10–5

4.1 × 101

Type M

Lung

5.4 × 10–7

2.0 × 100

3.3 × 10–5

1.2 × 102

Type S

Lung

6.4 × 10–7

2.3 × 100

8.4 × 10–6

3.1 × 101

Type M

Lung

5.5 × 10–7

2.0 × 100

2.3 × 10–6

8.5 × 100

Red marrow

3.2 × 10–9

1.2 × 10–2

Kidneys

3.5 × 10–8

1.3 × 10–1

1,600 y

αβγ

NS, IVC, U, F

Type M

Lung

4.7 × 10–7

1.7 × 100

2.2 × 10–6

8.1 × 100

39.3 d

βγ

NS, BC, U, F

Type M

Lung

8.9 × 10–9

3.3 × 10–2

1.8 × 10–9

6.6 × 10–3

NS, BC, U, F

Type S

Lung

1.0 × 10–8

3.7 × 10–2

2.1 × 10–9

7.7 × 10–3

NS, BC, U, F

Type M

Lung

3.3 × 10–8

1.2 × 10–1

1.7 × 10–8

6.3 × 10–2

Type S

Lung

3.9 × 10–8

1.4 × 10–1

3.4 × 10–8

1.3 × 10–1

NS, BC, U

Type M

Lung

3.2 × 10–9

1.2 × 10–2

6.8 × 10–10

2.5 × 10–3

U, F

Type F

Red marrow

4.6 × 10–9

1.7 × 10–2

3.0 × 10–8

1.1 × 10–1

374 d

βγ

46.5 h

βγ

28.8 y

β

58 / 3. COMPENDIUM OF RADIATION FACTS AND GUIDANCE

TABLE 3.13—(continued)

D

1.4 × 1010 y

U,D

2.5 × 10 y

232Th,

234

90Y

f

5

64.1 h

αβγ

αβγ

B

NS, IVC, U, F

NS, IVC, U, F

U

Type M

Lung

3.1 × 10–7

1.1 × 100

2.9 × 10–5

1.1 × 102

Type S

Lung

3.7 × 10–7

1.4 × 100

1.2 × 10–5

4.4 × 101

Type M

Lung

4.6 × 10

1.7 × 10

2.1 × 10

–6

7.8 × 100

Type S

Lung

5.4 × 10–7

2.0 × 100

6.8 × 10–6

2.5 × 101

Type M

Lung

4.5 × 10–9

1.7 × 10–2

1.6 × 10–9

5.9 × 10–3

–7

0

a

3.9 INTERNAL RADIONUCLIDE DEPOSITIONS

/ 59

Radionuclides are listed alphabetically by element. D is the possible presence of daughters with a half-life of <25 y (the radiations of the daughters are not included in the listing). bRadioactive half-life. c The primary radiations are listed. These include radiations emitted by dosimetrically-significant chain members. β = both positron and electron emission γ = includes conversion x-ray emissions as well as gamma rays EC = electron conversion n = neutrons dThe following symbols are used to indicate principal techniques for measuring external contamination or indicating internal exposure. The order of the symbols has no significance in the listing: BC = whole-body count (standard gamma detection methods) F = feces sample analyses IVC = special in vivo counting techniques useful for low-energy counting (wound monitoring, thyroid counting), or special low-energy x-ray or gamma detectors for chest counts (e.g., plutonium or americium counting) NS = nasal swab counted in laboratory if inhalation suspected U = urine sample analyses e Absorption type in the respiratory tract as defined in ICRP Publication 66 (ICRP, 1994a), Type F (fast), M (moderate), and S (slow. fUranium always comes as a mixture of the isotopes 238, 234 and 235. Natural uranium is composed of 99.3 % 238U, 0.711 % 235U, and 0.0058 % 234U by weight. In equilibrium, natural uranium has the same activity of 238U and 234U (48.9 %) and 2.2 % 235U. Enriched uranium is obtained when the concentration of 235U is increased to significantly >0.711 % by weight. When the concentration of 235U is decreased from 0.711 % by weight to 0.2 to 0.3 %, the material is called depleted uranium (Section 20.24).

60 / 3. COMPENDIUM OF RADIATION FACTS AND GUIDANCE absorbed doses to red marrow and lungs, with allowance for uncertainties typically involved in the dose estimates. CDGs for radioiodine are defined differently from those for other radionuclides because the cumulative dose to the thyroid is the pertinent measure of risk in this case, and FDA (2001) issued specific guidance regarding projected thyroid doses at which treatment for intake of radioiodine is indicated for different risk groups (Section 12). Based upon the recommendations and limits for emergency situations and knowledge of deterministic effects, the numerical values of dose used as a basis for computing the CDG intake values for different radionuclides, excluding isotopes of iodine, in adults are 0.25 Sv (25 rem) (50 y effective dose) for consideration of stochastic effects [based on the population-averaged nominal cancer fatality risk coefficient of 5 % Sv–1 derived from epidemiological data (ICRP, 2007), this represents about a 1.3 % lifetime risk of fatal cancer attributable to the radiation dose]; a 30 d RBEweighted absorbed-dose (see Terminology, Section 3.1.2) value of 0.25 Gy-Eq (25 rad-Eq) for consideration of deterministic effects to bone marrow (RBE = 2); and a 30 d RBE-weighted absorbed-dose value of 1 Gy-Eq (100 rad-Eq) for consideration of deterministic effects to the lungs (RBE = 7). CDG for an adult is the intake that satisfies the constraints on the effective dose and the 30 d absorbed doses to the red marrow and lungs. Table 3.14 provides a list of CDG values for nine noniodine radionuclides in adults. CDG values for additional noniodine radionuclides are given in Table 11.1. CDGs for children (0 to 18 y of age) and pregnant women for noniodine radionuclides are defined as one-fifth the adult value, reflecting the increased vulnerabilities during development and maturation (AAP, 2003). Children weighing >70 kg should be considered as adults. For an intake or expected intake of radioiodine, FDA recommends that KI be administered to adults >40 y of age if the projected dose to thyroid is ≥5 Gy (500 rad), to adults 18 to 40 y of age if the projected dose ≥0.1 Gy (10 rad), and to pregnant or lactating women or persons <18 y of age if the projected dose is ≥0.05 Gy (5 rad). In this Report, CDGs for 131I (the only isotope of iodine considered here) are derived separately for the following subgroups of the population, considering not only FDA dose guidelines for different risk groups but also projected differences with age in dose per unit intake of radioiodine (Section 20): adults of age >40 y; adults 18 to 40 y; pregnant or lactating women; and age groups 12 to 18, 7 to 12, 3 to 7, 0.5 to 3, and <0.5 y. The dose coefficient for thyroid (committed equivalent dose to thyroid per unit intake) for a reference adult is applied to each of the first three subgroups, and the coefficients for intake

TABLE 3.14—Abbreviated list of model predictions used to assess whether a radionuclide intake exceeds the CDG.a,b Values are for a reference adult (see Tables 11.1 and 11.2 for a full listing).c Effective Dose Radionuclide

Intake Moded

CDG (intake activity)

Formd Sv Bq–1

mrem μCi–1

Bq

μCi

Early Excretion and Retention (percentage of intake)

Early Excretion and Retention Levels Indicative of Intake of 1 CDG (dpm)e

Urinary Excretion 0 – 24 h

Retention in Chest at 24 hf

Total-Body Retention at 24 h

Urinary Excretion 0 – 24 h

Retention in Chest at 24 hf

Total-Body Retention at 24 h

Nasal Swab Soon After Inhalationg

5.6 × 107

4.3 × 108

4.4 × 107

Co

Inhalation

Type S

1.7 × 10–8

6.3 × 101

1.5 × 107

4.0 × 102

—h

6.4

49

—h

90Sr

Inhalation

Type F

3.0 × 10–8

1.1 × 102

8.3 × 106

2.3 × 102

6.8

NAi

49

3.4 × 107

NA

2.5 × 108

2.5 × 107

Inhalation

Type F

4.3 × 10–9

1.6 × 101

5.8 × 107

1.6 × 103

2.2

NA

58

7.7 × 107

NA

2.0 × 109

1.7 × 108

Ir

Inhalation

Type M

1.7 × 10–8 j

6.3 × 101 j

5.9 × 107

1.6 × 103

0.31

5.7

49

1.1 × 107

2.0 × 108

1.7 × 109

1.8 × 108

226Ra

Inhalation

Type M

2.2 × 10–6

8.1 × 103

1.1 × 105

3.1 × 100

0.16

5.8

50

1.1 × 104

4.0 × 105

3.4 × 106

3.4 × 105

238

Inhalation

Type S

6.8 × 10–6

2.5 × 104

3.7 × 104

9.9 × 10–1

0.07 l

6.4

49

1.5 × 103 l

1.4 × 105

1.1 × 106

1.1 × 105

Pu

Inhalation

Type M

3.1 × 10–5

1.1 × 105

8.1 × 103

2.2 × 10–1

0.021

5.80

50

1.0 × 102

2.8 × 104

2.4 × 105

2.4 × 104

239Pu

Inhalation

Type M

3.3 × 10–5

1.2 × 105

7.6 × 103

2.0 × 10–1

0.021

5.8

50

9.6 × 101

2.6 × 104

2.3 × 105

2.3 × 104

241

Inhalation

Type M

2.7 × 10–5

1.0 × 105

9.3 × 103

2.5 × 10–1

0.18

5.8

50

1.0 × 103

3.2 × 104

2.8 × 105

2.8 × 104

60

137

Cs

192

Uk

238

Am

aFor radionuclides other than isotopes of iodine, the CDG for a specific form of a radionuclide and mode of exposure is the intake activity estimated to result in the most restrictive of the following doses to an adult: a 50 y effective dose of 0.25 Sv (25 rem), an RBE-weighted 30 d absorbed dose to red marrow of 0.25 Gy-Eq (25 rad-Eq), or an RBE-weighted 30 d absorbed dose to lung of 1 Gy-Eq (100 rad-Eq). Fivefold lower CDGs are applied to children and pregnant women. The following alpha RBEs are applied: 20 for effective dose, two for 30 d RBE-weighted absorbed dose to red marrow, and seven for 30 d RBE-weighted absorbed dose to lungs. Effective dose is more restrictive than the 30 d RBE-weighted absorbed dose to red marrow or lung in most cases.

TABLE 3.14—(continued) b The following example illustrates how this table may be used. A patient enters the emergency room a few hours after an acute inhalation of 60Co, thought to be in the form of a relatively-insoluble oxide. External measurements indicate that total-body activity is <106 dpm and activity in the chest is <5 × 105 dpm. Measurements of urinary 60Co indicate that 24 h excretion is <104 dpm. These measurements are considerably lower than the reference 24 h retention and excretion values in this table corresponding to inhalation of 1 CDG of a relatively-insoluble form of 60Co (Type S). The measurements are also considerably lower than the reference values for inhalation of 60Co in moderately-soluble form (Type M) shown in Table 11.1. Thus, it appears that the patient has inhaled considerably <1 CDG, even if the inhaled material is somewhat more soluble than suspected. cFor application to children and pregnant women the CDG and the activities in urine, chest, total body, and nasal passages that correspond to 1 CDG should be divided by five. d If no information is available regarding the mode of intake or form of the radioactive material taken into the body, and if multiple cases are provided in this table for the radionuclide of concern, then measurement of activity in urine, chest, total body, or nasal swipe should be compared with the smallest listed value for urine, chest, total body, or nasal swipe, respectively, given for that radionuclide. e Divide by 60 to convert to becquerels and by 2.22 × 106 to convert to microcuries. f Retention in the chest refers to activity measured externally over the thoracic portion of the respiratory tract. This is assumed to represent activity in the lungs. gThe portion of inhaled activity found in a nasal swab in the early hours after inhalation of a radionuclide is highly variable, depending on such factors as aerosol size, the extent of nose breathing versus mouth breathing during the exposure, and the amount of nose blowing and wiping of the nostrils since the beginning of exposure. The listed activity for nasal swab represents 5 % of the inhaled amount, based upon experimental data and model predictions summarized in Section 10.3.1.1 and in greater detail in Section 16.4.1.2. The presence of radioactivity in a nasal swab is suggestive evidence of an inhalation exposure, particularly if both nares are contaminated. The absence of activity in a nasal swab does not establish that there was no inhalation exposure. h For these cases, calculation of an intake based on urinary excretion data is not recommended because of the high sensitivity of the estimate to the GI absorption fraction, which is not well established. Where feasible, decisions concerning treatment should be based on external measurement of activity in the chest, supplemented with measurement of activity in feces. Fecal excretion data can be interpreted on the basis of tabulations in Section 20. i NA = not applicable. In many cases of inhalation of radionuclides, external counts over the chest are not useful. For example, after inhalation of highly-soluble forms of radionuclides, activity quickly moves from the lungs to blood. Also, radionuclides that emit little if any penetrating radiation (e.g., the beta-emitter 3H or the alpha-emitter 210Po) are not detectable by external measurement. j The indicated dose is the RBE-weighted 30 d absorbed dose to the lungs, which is more restrictive than the effective dose in this case. kTable entries for 238U may also be applied to 234U or 235U. Chemical toxicity of uranium (nephrotoxicity) is generally of greater immediate concern than radiological toxicity following acute inhalation of elevated quantities of natural or depleted uranium. l Measurement of the urinary excretion rate should be supplemented with measurement of fecal excretion rate where feasible. A number of inhalation cases have been reported in which little or no activity was measured in urine for an extended period following significant exposure to an insoluble form of this radionuclide (see case studies for uranium and plutonium in Section 20).

3.10 RADIATION DOSE LIMITATION

/ 63

ages 15 y, 10 y, 5 y, 1 y, and 3 months are applied to age groups 12 to 18 y, 7 to 12 y, 3 to 7 y, 0.5 to 3 y, and <0.5 y, respectively. The CDG for radioiodine for a specific subgroup of the population is defined as FDA dose guideline value applicable to that subgroup, divided by the thyroid dose coefficient for that subgroup (Table 20.56). The resulting CDG intake values are given in Table 3.15 and also in Table 11.2. 3.9.2

Decorporation Therapy

Comprehensive information on decorporation therapy for internally-deposited radionuclides appears in Section 12. Tables 3.16 and 3.17 provide synopses of this information for quick access. 3.10 Radiation Dose Limitation The NCRP and ICRP dose limits in Tables 3.18 and 3.19 apply to planned, controllable, routine work, including small-scale and nonemergency contamination incidents. In the event of a major radiation emergency (e.g., a terrorist attack), it may be necessary to exceed these dose limits to perform lifesaving and other emergency-response activities as described in ICRP Publication 96 (ICRP, 2005a) and NCRP Report No. 138 (NCRP, 2001a). For lifesaving or equivalent purposes: workers may approach or exceed 0.5 Sv (50 rem) equivalent dose, 0.5 Gy (50 rad) absorbed dose for x-ray and gamma radiation to a large portion of the body and 5 Sv (500 rem) or 5 Gy (500 rad) for x and gamma radiation to the skin of the extremities (hands, feet, lower legs, and forearms) (NCRP, 1993). The decisive control for emergency responders working within or near the inner contamination area with personal protection equipment (PPE) is 0.5 Gy (50 rad) to the whole body. These are considered once-in-a-lifetime exposures (NCRP, 2005a). In Table 3.20, the Federal Emergency Management Agency (FEMA) dose limit is 0.25 Sv (25 rem) committed effective dose for lifesaving or protection of large populations when lower doses are not practicable; higher dose limits only on a voluntary basis to persons fully aware of the risks (FEMA, 2008).

Committed Equivalent Dose to Thyroid

CDG (intake activity)

Excretion and Retention During First 24 h (percentage of intake)

Excretion and Retention Levels During First 24 h Indicative of an Intake of 1 CDG (dpm)b

Group (Sv Bq–1)

(mrem μCi–1)

(Bq)

(μCi)

Urinary Excretion 0 – 24 h

Retention in Thyroid at 24 h

Total-Body Retention at 24 h

Urinary Excretion 0 – 24 h

Retention in Thyroid at 24 h

Total-Body Retention at 24 h

Adult >40 y

3.9 × 10–7

1.4 × 103

1.3 × 107

3.5 × 102

56

23

33

4.3 × 108

1.8 × 108

2.5 × 108

Adult 18 – 40 y

3.9 × 10–7

1.4 × 103

2.6 × 105

6.9 × 100

56

23

33

8.6 × 106

3.5 × 106

5.1 × 106

Pregnancy or lactation

3.9 × 10–7

1.4 × 103

1.3 × 105

3.5 × 100

56

23

33

4.3 × 106

1.8 × 106

2.5 × 106

Age 12 – 18 y

6.2 × 10–7

2.3 × 103

8.1 × 104

2.2 × 100

56

23

33

2.7 × 106

1.1 × 106

1.6 × 106

Age 7 – 12 y

9.5 × 10–7

3.5 × 103

5.3 × 104

1.4 × 100

56

23

33

1.8 × 106

7.3 × 105

1.0 × 106

Age 3 – 7 y

1.9 × 10–6

7.0 × 103

2.6 × 104

7.1 × 10–1

56

23

33

8.8 × 105

3.6 × 105

5.2 × 105

Age 0.5 – 3 y

3.2 × 10–6

1.2 × 104

1.6 × 104

4.2 × 10–1

56

22

32

5.3 × 105

2.1 × 105

3.0 × 105

Age <0.5 y

3.3 × 0–6

1.2 × 104

1.5 × 104

4.1 × 10–1

56

22

29

5.1 × 105

2.0 × 105

2.6 × 105

a The tabulated values are based on threshold doses estimated by FDA (2001) and listed in Table 12.14 for different risk groups, together with age-specific biokinetic and dose estimates for 131I inhaled as a vapor listed in Table 20.56 of this Report. bDivide by 60 to convert to becquerels and by 2.22 × 106 to convert to microcuries.

64 / 3. COMPENDIUM OF RADIATION FACTS AND GUIDANCE

TABLE 3.15—Model predictions used to assess whether an intake of 131I by inhalation as a vapor or ingestion exceeds the CDG.a

TABLE 3.16—Decorporation therapy recommendations for radionuclides of concern.a Radionuclides

Treatment

Preferred Prescription Consider DTPA DTPA BAL BAL Section 12.4.1 DTPA DMSA DMSA DTPA Section 12.4.1 Consider hydration and nonlabeled carbon DTPA Prussian blue DTPA DTPA Penicillamine DTPA DTPA DTPA

Aluminum hydroxide Penicillamine

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Consider DTPA DTPA British Anti-Lewisite (BAL), penicillamine BAL, dimercaptosuccinic acid (DMSA) Barium, calcium therapy (Section 12.4.1) DTPA BAL, Penicillamine, DMSA DMSA, DTPA, Ethylenediaminetetraacetic acid (EDTA) DTPA Barium, calcium therapy (Section 12.4.1) Consider hydration and nonlabeled carbon DTPA Prussian blue DTPA, EDTA (antacids are contraindicated) DMSA, DTPA, EDTA, N-acetyl-L-cysteine (NAC) EDTA, penicillamine, trientine DTPA DTPA DTPA Management depends on predominant isotopes present at time. Early: iodine; late: strontium, cesium, and others Aluminum hydroxide Consider penicillamine

3.10 RADIATION DOSE LIMITATION

Actinium Americium Antimony Arsenic Barium Berkelium Bismuth Cadmium Californium Calcium Carbon Cerium Cesium Chromium Cobalt Copper Curium Einsteinium Europium Fission products (mixed) Fluorine Gallium

Gold Indium Iodine Iridium Iron Lanthanum Lead Manganese Magnesium Mercury Molybdenum Neptunium Nickel Niobium Palladium Phosphorus Plutonium Polonium Potassium Promethium Radium

Treatment BAL, penicillamine DTPA KI, consider saturated solution of potassium iodide (SSKI), propylthiouracil, methimazole or potassiumiodate Consider DTPA, EDTA Deferoxamine (DFOA), deferasirox, DTPA, DFOA and DTPA together DTPA DMSA, EDTA, EDTA with BAL DFOA, DTPA, EDTA Consider strontium therapy (Section 12.4.5) BAL; EDTA; penicillamine; DMSA Limited clinical experience Consider DFOA and/or DTPA BAL, EDTA DTPA Penicillamine, DTPA Phosphorus therapy (Section 12.4.4) DTPA, DFOA, EDTA, DTPA and DFOA together BAL, DMSA, penicillamine Diuretics DTPA Radium, strontium therapy (Section 12.4.5)

Preferred Prescription BAL DTPA KI Consider DTPA DFOA DTPA DMSA DTPA Consider strontium therapy BAL Consider DFOA and/or DTPA BAL DTPA Penicillamine Phosphorus therapy DTPA BAL Diuretics DTPA Section 12.4.5

66 / 3. COMPENDIUM OF RADIATION FACTS AND GUIDANCE

TABLE 3.16—(continued) Radionuclides

3.10 RADIATION DOSE LIMITATION

Rubidium Prussian blue Prussian blue Ruthenium DTPA, EDTA DTPA Scandium DTPA DTPA Silver No specific therapy. Sodium Diuretic and isotopic dilution with 0.9 % NaCl Diuretic and isotopic dilution with 0.9 % NaCl Strontium Radium, strontium therapy (Section 12.4.5) Section 12.4.5 Sulfur Consider sodium thiosulfate Consider thiosulfate Technetium Potassium perchlorate Potassium perchlorate Thallium Prussian blue Prussian blue Thorium Consider DTPA Consider DTPA Force fluids Water diuresis Tritium (3H) Uranium Bicarbonate to alkalinize the urine. Consider dialysis Bicarbonate Yttrium DTPA, EDTA DTPA Zinc DTPA, EDTA, zinc sulfate as a diluting agent. DTPA Zirconium DTPA, EDTA DTPA aThe majority of these drugs are not approved by FDA for the indications listed in this table (see Section 12 for further information and warnings).

/ 67

Dosage

Acetylcysteine [N-acetyl-L-cysteine (NAC)] (Section 12.4.2)

Consider dosage as for acetaminophen overdosage, start at 140 mg kg –1 oral loading dose (RxList, 2009).

Deferoxamine (DFOA) mesylate (Section 12.3.1)

FDA does not specify age: DFOA mesylate injectable; IM is preferred. 1 g IM or IV (2 ampules) slowly (15 mg kg –1 h–1); Repeat as indicated as 500 mg IM or IV q4h × 2 doses; then 500 mg IM or IV every 12 h for 3 d.

Dimercaprol [British Anti-Lewisite (BAL)] (Section 12.3.2)

FDA does not specify age: IM: 300 mg per vial for deep IM use, 2.5 mg kg –1 (or less) every 4 h for 2 d, then twice daily for 1 d then daily for days 5 to 10.

Diethylenetriaminepentaacetate (DTPA, calcium or zinc) (pentetate calcium trisodium and Pentetate zinc trisodium) (Section 12.3.3)

Adults: IV: 1 g in 5 mL IV push over 3 to 4 min or IV infusion over 30 min diluted in 250 mL of 5 % dextrose in water, Ringers lactate or normal saline. Nebulized inhalation: 1g in 1:1 dilution with sterile water or normal saline. Children under 12 y: 14 mg kg –1 IV as above, not to exceed 1 g.

Edetate calcium disodium [Ethylenediaminetetraacetic acid (EDTA)] Section 12.3.4)

FDA does not specify age: Ca-EDTA (edetate calcium disodium); 1,000 mg m–2 d–1 added to 500 mL 5 % dextrose or 0.9 % sodium chloride infused over 8 to 12 h. This same dosage can be given IM divided into equal doses spaces 8 to 12 h apart.

Penicillamine (Section 12.3.5)

FDA does not specify age: Oral: 250 mg daily between meals and at bedtime. May increase to 4 or 5 g daily in divided doses.

Phosphorus therapy Potassium phosphate, dibasic (Section 12.4.4)

Oral: 250 mg phosphorus per tablet. Adults: 1 – 2 tablets oral four times daily with full glass of water each time, with meals and at bedtime. Children >4 y of age: 1 tab oral four times daily.

Potassium iodide (KI) (Section 12.4.3)

Oral: tablets or liquid. Drug dose varies between 16 and 130 mg daily depending on age, thyroid exposure level, and whether or not pregnant or lactating (Table 12.14).

68 / 3. COMPENDIUM OF RADIATION FACTS AND GUIDANCE

TABLE 3.17—Dose schedules for drug or treatment modalities. Drug or Treatment Modality

Oral: 50 mg tablets, 2 tablets three times daily for 8 d. FDA does not specify age.

Prussian blue (Section 12.3.6)

Oral: Adults and adolescents 3 g three times daily. Children 2 to 12 y of age: 1 g three times daily.

Sodium bicarbonate (for uranium only) (Section 12.4.7)

Oral or IV (Table 12.22).

Radium and strontium therapy (Section 12.4.5)

Section 12.4.5.

Succimer [dimercaptosuccinic acid (DMSA)] [Chemet® (Schwarz Pharma, Monheim, Germany)] (Section 12.3.7)

FDA approved pediatric dosing: Start dosage at 10 mg kg –1 or 350 mg m–2 oral every 8 h for 5 d. Reduce frequency of administration to 10 mg kg –1 or 350 mg m–2 every 12 h (two-thirds of initial daily dosage) for an additional two weeks of therapy. A course of treatment lasts 19 d.

Water diuresis (Section 12.4.6)

Oral: Fluids >3 – 4 L d–1.

aUnless b

noted otherwise, the references for these dose schedules are given in the listed sections. Dosage notations: IV = intravenous injection bid = twice per day IM = intramuscular injection tid = three times per day mEq = milliequivalent qid = four times per day PO = per os or orally qd = every day q12h = every 12 h q4h = every 4 h

3.10 RADIATION DOSE LIMITATION

Propylthiouracil (Section 12.4.3)

/ 69

70 / 3. COMPENDIUM OF RADIATION FACTS AND GUIDANCE TABLE 3.18—NCRP dose-limit recommendations (NCRP, 1993; 2005a).a Population

Dose Limit

Occupational Effective dose, annual

50 mSv (5 rem)

Effective dose, cumulative

10 mSv (1 rem) × age

Equivalent dose, annual for tissues and organs; lens of eye

150 mSv (15 rem)

Skin, hands, and feet

500 mSv (50 rem)

Public Effective annual (frequent or continuous) (infrequent)

1 mSv (0.1 rem) 5 mSv (0.5 rem)

Equivalent dose, annual for tissues and organs; lens of eye Skin, hands, and feet aDose

15 mSv (1.5 rem) 50 mSv (5 rem)

limits, both NCRP and ICRP, apply only to planned exposure situations.

TABLE 3.19—ICRP individual dose limits (ICRP, 2007).a Type

Dose Limit

Public Exposure Individual

1 mSv (100 mrem) effective dose in a year

Lens of eye

15 mSv (1.5 rem) annaul equivalent dose

Skin

50 mSv (5 rem) annaul equivalent dose

Occupational Exposure Individual worker

20 mSv (2 rem) effective dose average over 5y

Lens of eye

150 mSv (15 rem) annaul equivalent dose

Skin

500 mSv (50 rem) annaul equivalent dose

Hands and feet

500 mSv (50 rem) annaul equivalent dose

a Recommendations of ICRP do not apply directly to persons in the United States, but may apply to those in other nations.

TABLE 3.20—Emergency worker guidelines in the early phase (FEMA, 2008).a Total Effective Dose Equivalentb Guideline

Activity

Condition

All occupational exposures

• All reasonably achievable actions have been taken to minimize dose.

0.1 Sv

Protecting valuable property necessary for public welfare (e.g., a power plant).

• Responders have been fully informed of the risks of exposures they may experience. • Dose >0.05 Sv (5 rem) is on a voluntary basis. • Appropriate respiratory protection and other personal protection is provided and used. • Monitoring available to project or measure dose.

Lifesaving or protection of large populations. It is unlikely that doses would reach this level in a radiological dispersal device incident. However, worker doses >0.25 Sv (25 rem) are conceivable in a catastrophic incident such as an improvised nuclear device incident.

• All appropriate actions and controls have been implemented; however, >0.05 Sv (5 rem) is unavoidable. • Responders have been fully informed of the risks of exposures they may experience. • Dose >0.05 Sv (5 rem) is on a voluntary basis. • Appropriate respiratory protection and other personal protection is provided and used. • Monitoring available to project or measure dose.

(10 rem)

0.25 Sv (25 rem)c

/ 71

aIn the intermediate and late phases, standard worker protections, including the 0.05 Sv (5 rem) occupational dose limit, would normally apply. b The projected sum of the effective dose equivalent from external radiation exposure and committed effective dose equivalent from internal radiation exposure. cEPA (1992) states that “Situations may also rarely occur in which a dose >0.25 Sv (25 rem) for emergency exposure would be unavoidable in order to carry out a lifesaving operation or avoid extensive exposure of large populations.” Similarly, NCRP and ICRP raise the possibility that emergency responders might receive an equivalent dose that approaches or >0.5 Sv (50 rem) to a large portion of the body in a short time (NCRP, 1993). If lifesaving emergency-responder doses approach or >0.5 Sv (50 rem), emergency responders must be made fully aware of both the early and the late (cancer) risks of such exposure.

3.10 RADIATION DOSE LIMITATION

0.05 Sv (5 rem)

4. Radiation-Safety Guidance for First Responders 4.1 Introduction Medical and radiation-safety personnel who are first responders to an incident in which persons may have been exposed to radionuclides have six major objectives: • • • • •

provide medical aid to exposed individuals; identify irradiated and contaminated individuals; detect and identify radioactive material; identify sources of external radiation; control the radionuclide contamination, preventing the spread of radionuclides beyond the incident site; and • initiate decontamination of individuals and the site. The highest priority should be to provide medical care to all injured, exposed and unexposed. However, in principle, all of these objectives should be pursued simultaneously by all first responders. It is important that these major objectives be achieved with the utmost attention to protection of exposed persons as well as the professionals attending them. This section describes the steps to be taken to achieve this protection. An effective response to an incident in which persons may be contaminated with radionuclides, whether the incident is small involving one or a few individuals or very large involving large numbers of individuals, requires that qualified experts, medical personnel (physicians, nurses, and medical technicians) and radiationsafety personnel (radiation safety officers, health physicists, and radiation protection technicians) work as a team. As described in Section 18, medical and radiation-safety personnel will also provide guidance to and cooperate with other first responders such as fire and law-enforcement officials. This is emphasized in the following guidance. In an ideal situation, a radionuclide contamination incident will occur in a location where both medical and radiation-safety staff are 72

4.2 GENERAL INSTRUCTIONS FOR FIRST RESPONDERS

/ 73

available. Fortunately that is almost always the case since nearly all incidents are accidents occurring in facilities routinely handling radioactive materials, such as hospitals, research laboratories, universities, government nuclear sites, nuclear power stations, and industries using radionuclides. These institutions generally employ radiation-safety officers and have trained medical staff available. When contamination incidents occur, extensive prehospital care is generally possible, depending upon the training of the personnel and the availability of instrumentation. Ideally, the personnel at the onsite facility will have removed all transferable radioactive material from the patient, estimated the severity of internal contamination, and provided emergency first aid for wounds before the patient is moved to the hospital. In general, the hospital is used only for definitive medical care. While the following guidance is more specific for facilities having radiation-safety and medical support, the principles can be applied to nearly all radionuclide contamination situations. Obviously, when large incidents occur and many people are exposed, considerable ingenuity is required to manage contaminated individuals efficiently and effectively. 4.2 General Instructions for First Responders The radiological nature of an incident may not be immediately obvious, especially in the event of a large explosion that causes confusion, ignites fires, damages structures, and injures and kills bystanders. Until the radiological nature of an incident is recognized (and, to some extent, even afterwards), the highest priority should be devoted to rescue and lifesaving operations, performing triage on injured persons, evacuating the most seriously injured, and other immediately necessary actions (e.g., firefighting). Once the radiological nature of the incident is recognized, it will also be important to determine the nature and extent of the contamination, after which the entire contaminated area should be cordoned off and radiation warning signs posted and radiation control zones established. Inner and outer contaminated areas and a secured area should be established as described in Sections 4.3.3 and 18. The potential offsite transport of radioactive materials through air or water contamination and by people and vehicles passing through contaminated areas will be the concern of those responsible for public health and environmental safety (Section 18). The presence of radioactive contamination will determine the need for PPE such as gloves, respiratory protection, and shoe covers for those entering the area. Persons leaving a contaminated

74 / 4. RADIATION-SAFETY GUIDANCE FOR FIRST RESPONDERS area must remove their PPE and (if necessary) decontaminate themselves prior to exit (Section 5.2). The presence of elevated radiation levels will determine the radiological stay-times for persons working in the area. U.S. Nuclear Regulatory Commission (NRC) regulations such as 10 CFR 20.1003 (NRC, 2002a) and 10 CFR 20.1601 (NRC, 2002b) define and specify postings and controls for these areas. Radiologically-controlled areas will be established to recognize both contamination and radiation levels. Personnel responding to a radiological incident will use radiation detectors to determine the location of perimeter boundaries. Radiation dose-rate measurements can be read directly from the meters in units of mGy h–1 or mR h–1. Contamination limits are typically provided in units of becquerel or disintegrations per minute in a reference area3 [e.g., becquerel per square centimeter (Bq cm–2), or disintegrations per minute per 100 square centimeter (disintegrations per minute 100 cm–2)]. Radiation detectors do not read directly in units of becquerel or disintegrations per minute; they read in counts per minute. Each meter has a counting efficiency for each energy and type of radiation it is measuring; the meter reading is equal to the amount of contamination multiplied by the counting efficiency, which is calculated when a meter is calibrated. To convert from the meter reading of counts per minute to the required units of disintegrations per minute, the user must divide the displayed count by the counting efficiency. For example, if a reading of 1,000 cpm is displayed on the meter face, and the meter is known to have a counting efficiency of 10 %, then the amount of contamination present is equal to 1,000 cpm divided by 0.10 = 10,000 dpm. 4.3 Guidance for First Responders NCRP Commentary No. 19, Key Elements of Preparing Emergency Responders for Nuclear and Radiological Terrorism (NCRP, 2005a), provides emergency responders sound advice for conducting their critical work in a dangerous radiation environment. Although the focus of Commentary No. 19 is nuclear and radiological terrorism, the advice is applicable to the full range of potential radionuclide exposure situations described in Section 17. Emergency responders should have a basic understanding of external radiation exposure, external and internal radioactive contamination and the medical consequences of each. They should understand 3

1 Bq is 1 disintegration per second, so 1 Bq = 60 dpm.

4.3 GUIDANCE FOR FIRST RESPONDERS

/ 75

what radionuclide contamination is and what processes are necessary to mitigate contamination and alleviate potential medical consequences, and know the appropriate measures to protect themselves, others at the site, and members of the public. 4.3.1

First on the Scene

NCRP Commentary No. 19 (NCRP, 2005a) provides recommendations for a radiation emergency until designated authorities declare the emergency over. At that time, the established radiation protection procedures for both occupational and public exposure would normally be reinstituted, as well as any special arrangements for long-term control of a continuing elevated radiation environment. Once a response to a radiological incident is initiated, the first emergency responders to a scene should be assigned radiationmonitoring equipment to record exposure rate and cumulative dose. The instrument should be designed so the emergency responder (e.g., a bunker gear-clad firefighter) can readily interpret the reading and operate the controls. The radiation-monitoring instrument for emergency responders should provide a digital readout for exposure rate, cumulative dose and estimated stay-time at the current exposure rate. The readout should be linked to a visual indication such as green, yellow and red indicator lights because of likely complex noise and activity environment at the response scene (NCRP, 2005a). While the fundamental concept of keeping all radiation exposures as low as reasonably achievable (the ALARA principle) applies, it may not be realistic to apply other traditional radiation protection guidelines for limitation of radiation dose. The traditional guidelines are based on an assumption of low-level chronic exposure over long periods, and govern activities that are more controllable than the release of radionuclides in a work place, or in a nuclear or radiological terrorism incident (NCRP, 2005a). 4.3.2

Immediate Goals for Protection of Exposed Individuals

The two important goals of radiation protection in such an emergency are: • to prevent immediate injuries and deaths due to acute highlevel radiation exposure including intakes of radionuclides, and • minimize long-term effects (i.e., cancer) associated with lower levels of radiation exposure from both external and internal sources.

76 / 4. RADIATION-SAFETY GUIDANCE FOR FIRST RESPONDERS Minimizing radiation doses in both cases, by taking advantage of the basic features of radiation protection (i.e., increasing the distance from the source, limiting time of exposure, utilizing intervening shielding, removing individuals from contaminated areas, and controlling contamination including external decontamination of individuals), is much more effective than subsequent medical treatment and countermeasures. The main protective actions in a radiation incident will be medical treatment for those injured by an explosion, evacuation from the affected area, external decontamination of those who were contaminated, and assessment of external and internal contaminations. 4.3.3

Control Areas

The area in which the release of radionuclides occurs should be isolated to prevent the spread of contamination to unaffected areas. Any emergency-response personnel entering the area should be prepared to work in a contaminated environment, and those leaving the area should be monitored and decontaminated as necessary. To prevent the spread of radionuclides beyond the area of the incident, control areas should be established. These are described further in Section 18. Figure 4.1 provides a generic example of how control areas might be established. Three defined areas are shown: • inner contaminated area where the radionuclide release occurred; • outer contaminated area where released activity may be transported such as by explosions, air currents or inadvertently by people walking and vehicles driving from the release area; and • secured (clean) area, where entry and egress are controlled to minimize further contamination of people, facilities and the environment. This radionuclide control concept can apply to a broad spectrum of accidental and deliberate contamination incidents such as explosions, fires, ruptures of sources, and spills of radioactive materials in industrial settings, laboratories and hospitals. Section 5.3.4 describes in detail the operation of radiation and radionuclide controlled areas, whether at the site of a contamination incident or at a hospital or other facility where contaminated patients are being examined, decontaminated or treated. The particular situation will determine the configuration of control zones and the criteria for establishing contamination levels. In situations where the released radionuclides are totally contained at the site of release, the outer contaminated area would not be

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Fig. 4.1. A generic example of areas designated for specific activities in management of exposed persons after an incident involving the release of radionuclides. Several areas would be appropriately cordoned off and identified (adapted from IAEA, 2005b).

needed. Control areas or zones could be defined quite differently, depending upon the nature and the magnitude of the radionuclide source. For example, to attain the same objective, control of the released activity, an incident in a building would require an approach different from those in rural or urban sites. Applying the example in Figure 4.1 to a spill in a laboratory, the inner contaminated area is the location of a radionuclide release or spill (e.g., on a bench or laboratory floor). A radiation control point would be immediately established (perhaps only a step-off pad) to minimize the spread of the contamination. The laboratory or the corridor leading to the laboratory might be established as the outer contaminated area where activity may have, or is likely to, spread as a result of movement of people, equipment or ambient air. The whole building might be defined as the third, secured (clean) area, where entry and exit of persons and equipment would be controlled. The controlling radiation dose rates and contamination levels established by the radiation-safety personnel would depend upon the nature of the incident. These control areas should be cordoned off with barriers in place or locked doors and identified appropriately. Applying this configuration to a high-level radiological terrorist incident, the site might be segmented into radiation control zones as described in NCRP Commentary No. 19 (NCRP, 2005a) (Section 18.3).

78 / 4. RADIATION-SAFETY GUIDANCE FOR FIRST RESPONDERS In the generic example, Figure 4.1, triage, medical-response, and decontamination activities are located outside the perimeter of the outer contaminated area but within the perimeter of the secured area. This concept of control zones is used in general response procedures for hazardous materials. Control areas at hazardous materials incident sites are designated based upon safety and the degree of hazard. In radionuclide contamination incidents control zones help to limit the absorbed doses received by individual emergency responders as well as to facilitate effective management of exposed individuals. Appendix C addresses risks to first responders. Section 18 gives further descriptions of activities performed in each area. A command post might be outside the perimeter of the controlled area, as well as other support functions deemed necessary to manage the incident. The inner contaminated area, that area immediately surrounding the release of a radionuclide, should extend as far as necessary to prevent individuals from receiving radiation doses sufficient to cause early radiation effects (IAEA, 2003). A prudent approach is to assume the activity present in the inner contaminated area has the potential to produce adverse health effects and thus those persons in the area should be restricted to time-sensitive, mission-critical activities such as lifesaving. As noted before, the outer contaminated area may not be necessary if significant quantities of activity have not been spread from the inner area. Within the inner contaminated area, the appropriate actions are to evacuate or temporarily relocate the people, isolate the area and ensure that all emergency workers inside the area follow appropriate personal protection guidelines. More specific guidance on working in radiologically-controlled areas is given in Sections 5.2 and 5.3. 4.3.4

Protection of First Responders

NCRP recommends that for lifesaving or equivalent purposes, emergency workers may approach or exceed 0.5 Sv (50 rem) equivalent dose, 0.5 Gy (50 rad) absorbed dose, for x-rays and gamma radiation to a large portion of the body and an equivalent dose of 5 Sv (500 rem) to the skin (NCRP, 1993; 2001a). These are considered once-in-a-lifetime exposures. The decisive control for emergency responders working within or near the inner contaminated area with PPE is the 0.5 Gy (50 rad) whole-body value. ICRP recommends doses to emergency responders involved in lifesaving procedures be kept <1 Sv (100 rem) to avoid deterministic health effects. For those involved in recovery operations, normal occupational limits apply, effective dose of 50 mSv (5 rem) (ICRP, 2005a).

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Standard PPE including respiratory protection devices should be sufficient for most emergency responders conducting lifesaving and other critical missions (Section 5 describes PPE). In a major incident during the initial scoping of the response, radiation levels should be communicated to the incident commander or other responsible individual(s), who must assess the “life hazard” to exposed persons and the emergency responders. Additional equipment is required to screen for contamination of large numbers of people and for screening at emergency facilities (NCRP, 2005a). In large-scale releases of radiation and radioactive materials, individual pocket-sized alarming dosimeters should be used by all emergency responders in the inner contaminated area and, if available, by responders in the outer contaminated area (NCRP, 2005a). The radiation-monitoring instrument for emergency responders should provide a digital readout for exposure rate, cumulative dose, and estimated stay-time at the current exposure rate. The instrument would warn the wearer when the radiation exposure approaches a predetermined level or is likely to reach the level. The instrument, which would also serve as a dosimeter, should produce a warning, such as a steady red indicator (and perhaps also an audible indication), that the wearer has reached a cumulative absorbed dose of 0.5 Gy (50 rad). The instrument would display a blinking yellow light when the exposure rate is >0.1 mGy h–1 (10 mR h–1) but <0.1 Gy h–1 (10 R h–1) and would indicate the presence of significant radiation by displaying a blinking red light at exposure rates >0.1 Gy h–1 (10 R h–1). The instrument would display a steady green light when the exposure rate is <0.1 mGy h–1 (10 mR h–1). This combination of digital and visual readout would warn the wearer of the radiation levels while providing enough information to manage the exposure and to complete important tasks (NCRP, 2005a). The cumulative absorbed dose that triggers a decision on whether to remove an emergency responder from the inner contaminated area is 0.5 Gy (50 rad). The cumulative absorbed dose received by an emergency responder while working in or near the inner contaminated area should be recorded.

5. Performing Surveys and Controlling Personnel and Area Contamination While the following is primarily for those who may not be trained radiation-safety personnel, it may also be helpful to those who have not had recent experience with radiation detection instrumentation and radionuclide contamination. In all cases, to ensure proper use of radiation detection instrumentation, consultation with knowledgeable experts should occur whenever possible (see Section 19 for information on radiation detection instrumentation). Additional information is available in NCRP Commentary No. 19 (NCRP, 2005a) and in Background Information on FEMA-REP-22: Contamination Monitoring Guidance for Portable Instruments Used for Radiological Emergency Response to Nuclear Power Plant Accidents (FEMA, 2002). It is essential that PPE be used and worn correctly. Therefore, whenever possible it should be under the direction of trained experts. 5.1 Contamination Surveys First responders may be required to perform various surveys to detect external radiation sources and radioactive contamination. A major objective of onsite medical assessment or triage (Section 6) is an early identification of persons exposed to external radiation and those externally and internally contaminated with radionuclides. Other stages in the management of contaminated individuals onsite and at treatment facilities, require further and more thorough surveys for radionuclide contamination. The following briefly describes, in general, the steps to be taken in operating various radiation survey instruments to perform surveys of persons who may have been exposed, their clothing, air samples, nasal swabs, and other items of interest, as well as decontamination areas, emergency facilities, hospitals, vehicles used to transport contaminated persons, equipment and surveys of the professionals responding to contamination incidents. Instrumentation used in these types of surveys is described in detail in Sections 19.2.4 through 19.2.7.3. Additional information on these 80

5.1 CONTAMINATION SURVEYS

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types of instruments and their applications is available in references such as: Radiation Detection and Measurement (Knoll, 2000), Background Information on FEMA-REP-22: Contamination Monitoring Guidance for Portable Instruments Used for Radiological Emergency Response to Nuclear Power Plant Accidents (FEMA, 2002), Direct Determination of the Body Content of Radionuclides (ICRU, 2003), and Introduction to Health Physics (Cember and Johnson, 2008). 5.1.1

How to Perform Surveys of Individuals, Clothing, Samples and Surfaces

It is advisable that these surveys be performed by trained experts or at least under their guidance whenever possible. 1.

2.

3.

4.

5.

Check the test or calibration certificate. Radiation survey instruments will generally be accompanied by a test or calibration certificate. The information on this certificate should be documented with the results of the surveys taken with the instruments. Turn on the meter, check the battery charge, observe the meter for response (many meters have a check source attached, or one should be obtained), switch to highest setting, engage audible monitor, if available, and be aware of how different scales on the meter may indicate different exposure rates. Obtain a background reading at ~1 m from the contaminated surface. When possible, the background should be taken in the general area, but not near the incident site. Subtract background from survey readings. Hold detector <1 cm (0.5 inch) for beta/gamma, 0.25 inch for alpha from the item being surveyed (avoid touching the contaminated material) and move it at ~3 to 5 cm (one to two inches) s–1. Turn switch to lower scales until the meter reading is less than three-quarters of the full scale. Instrumental contamination surveys cannot be performed if background level is too high (e.g., 300 cpm on a frisker)4 (Section 19.2.7.2). In these cases, smears must be used and counted in a low background area. Do not take smear samples from contaminated individuals in a high background area; take the individuals to a lower background area.

4Generic

name for hand-held instrument generally used for surveying external surfaces of people and objects.

82 / 5. PERFORMING SURVEYS AND CONTROLLING CONTAMINATION 6.

7. 5.1.2 1.

2. 3.

5.1.3 1.

2. 3.

4. 5. 6.

If the detector gives results in count rate, use the instrument’s calibration factor to convert measurements into activity Bq cm–2 (dpm cm–2) or, if the probe size is 100 cm2 (the size of newer probes), record the results as activity per 100 cm2 Record results on a survey map or diagram and note areas with high contamination levels. Observe if results could have been influenced by uneven surfaces or from visible contamination (especially important for alpha or low-energy beta contamination). Document results (Appendix A). How to Perform a Beta/Gamma-Radiation Area Survey Turn on the meter, check battery charge, observe the meter for response (many meters have a check source attached, or one should be found), engage audible monitor, if available, and be aware of how different scales on the meter may indicate different exposure rates. Hold detector or meter about waist height and walk slowly through area. Note areas with elevated readings on survey maps or diagrams. How to Perform an Alpha-Radiation Survey of Contaminated Areas, Individuals and Samples Turn on the meter, check battery charge, observe the meter for response (many meters have a check source attached, or one should be found), engage audible monitor, if available, be aware of how different scales on the meter may indicate different exposure rates. If the meter has a protective plastic cover over the alpha-sensitive area, this must be removed prior to performing the survey. Pass detector (probe) very close to, but not touching, suspected areas of contamination at the incident site. Pass detector (probe) very close to, but not touching, air samples, smears, and nasal swabs, and suspected areas of contamination on individuals, clothing and equipment. Identify areas with elevated readings on survey maps or diagrams. Document external alpha contamination of individuals on a body diagram (Appendix A). Document survey results of air samples, smears, nasal swabs, clothing, and other items relevant to external and internal contamination assessments.

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General characteristics of different survey instruments and different types of on- and offsite surveys are given in Tables 5.1 and 5.2. 5.2 Personal Protection Equipment It is important that personal protection equipment (PPE) be used or worn correctly to ensure adequate protection of the individual. Respiratory protection is especially sensitive to being correctly fitted to the face as well as being the appropriate type for the situation. Therefore, it is advisable that the use of PPE be under the direction of trained experts. 5.2.1

Examples of Personal Protection Equipment

The following items of protective equipment should be used as necessary, according to the risk and level of contamination: • • • • • • • 5.2.2

gloves shoe covers and booties coveralls outer shoe covers head/hair covering tape to close open ends of clothing respiratory protection Personal Protection Equipment Inspection

Prior to dressing, PPE should be inspected as follows: 1. 2. 3. 4.

5.

examine gloves, shoe covers, and coveralls for rips, tears or split seams; examine gloves for holes or split seams; examine shoe covers for holes and for proper size; verify proper operation of respiratory protection: a. ensure nuisance masks or N95 filters5 do not have holes; b. ensure all masks fit tightly around nose and chin; c. confirm sufficient air supply in forced-air masks; and d. examine all hoses for damage and leaks; remove watches, jewelry, rings, etc., to reduce the risk of making holes in the PPE.

5By NIOSH classification an N95 mask or filter traps 95 % of particles with a diameter of 0.3 μm or larger. Any mask, including an N95 mask, should be tested to ensure a good fit prior to wearing.

Meter

Probe

Record Results in:a

Use for These Surveys

General survey instrument

GM

Becquerel (dpm)

Contamination

Hand and foot monitor

GM

Becquerel (dpm)

Contamination on hands and feet when leaving a room

General survey instrument

Energy-compensated GM

mGy or mSv h–1 (mR h–1) becquerel (dpm)

Radiation levels Contamination

General survey instrument

NaI scintillation

mGy or mSv h–1 (mR h–1) becquerel (dpm)

Radiation levels Contamination

Ion chamber or micro-R meter

Ion chamber or micro-R meter

mGy or mSv h–1 (mR h–1) or μGy h–1 (μR h–1)

Radiation levels

Alpha survey instrument

Zinc sulfide

Becquerel (dpm)

Contamination

Portable spectroscopy survey meter

Germanium semiconductor

Output varies; radionuclide and activity determination

Radionuclide identification and possible quantification

aMany

instruments readout in counts per minute. These readings should be converted to becquerel and or disintegrations per minute using the calibration factors appropriate for the particular instrument.

84 / 5. PERFORMING SURVERYS AND CONTROLLING CONTAMIANTION

TABLE 5.1—Radiation instrumentation used for contamination and radiation surveys.

TABLE 5.2—Types of surveys and appropriate instruments. Survey

Survey Type

Meter and Probe

Record Results in:

Contamination

Hand and foot monitor (GM)

Becquerel (dpm)

Smear wipe

Removable contamination

Count wipes in well counter, or with meter

Becquerel (dpm)

Area survey

Radiation

Energy-compensated GM, ion chamber or micro-R meter

mGy or mSv h–1 (mR h–1) or μGy or μSv h–1 (μR h–1)

Spills, personnel surveys

Contamination

GM and/or alpha detector

Becquerel (dpm)

Radionuclide identification

Contamination

Portable spectroscopy survey meter

Output varies, radionuclide and activity determination

Personnel

Contamination

Portal monitor

Becquerel (dpm)

5.2 PERSONAL PROTECTION EQUIPMENT

Hands and feet (exiting room)

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86 / 5. PERFORMING SURVEYS AND CONTROLLING CONTAMINATION 5.2.3

Dressing in Personal Protection Equipment

The equipment should be put on in this order. 1. 2. 3. 4. 5. 6.

turnout gear or coveralls; shoe covers (tape at ankles if necessary and possible), wear two pair or booties over rough terrain; gloves (tape inner gloves at wrists if necessary); respiratory protection (if advisable); hood or head covering (if applicable). Tape hood to respiratory protection; and dosimeter (should be easily accessible)

Table 5.3 provides important information on appropriate PPE to be used during different phases of response to an incident. • if time and conditions permit, tape coveralls or turnout gear at wrists and ankles; • if there are not enough personal dosimeters for each person, give a dosimeter to at least one person in each response group (e.g., fire-hose team, rescue party).; • radioactive contamination on the ground or on an individual is only very rarely hazardous; • lifesaving actions (rescuing people, firefighting, stabilizing structures) may be performed, provided responders wear appropriate PPE and respiratory protection, even when contamination levels are very high; and • perform radiation dose rate and contamination level surveys at the earliest opportunity. 5.2.4

Removing Personal Protection Equipment

Remove protective equipment in this order: 1. 2. 3. 4. 5. 6. 7. 8.

outer gloves give dosimeter to health and safety personnel tape at ankles, wrists coveralls (coat and pants) head cover, helmet and or hoods respiratory protection shoe covers (step to clean area) inner gloves

TABLE 5.3—PPE for emergency and recovery phases. Incident

• Hot particles on skin • Inhaling activity • Swallowing radioactive materials

Appropriate PPE Emergency Phase

• Turnout gear (gloves, boots, coat) • Full-face air-purifying respirator, forced air, or preferred, a self-contained breathing apparatus • Personal dosimeter

Recovery Phase

• Coveralls, outer clothes • Gloves (rubber, leather, surgical) • Shoe covers • Nuisance mask for dust (N95) unless air sampling indicates otherwise • Personal dosimeter

5.2 PERSONAL PROTECTION EQUIPMENT

Area contamination with radioactive dusts, liquids, or airborne spray of activity

Radiological Concerns

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88 / 5. PERFORMING SURVEYS AND CONTROLLING CONTAMINATION 5.2.5 1. 2.

3.

4. 5.

Actions to be Taken after Personal Protection Equipment is Removed Perform whole-body contamination survey in accordance with Section 7 of this Report. As it is removed, survey and dispose of contaminated PPE or set it aside for decontamination as warranted. Disposable PPE is usually deposited in waste containers without survey. Place contaminated PPE and clothing into designated radioactive-waste receptacles. If required, package each person’s removed PPE in a separate bag before placing into waste receptacle. Survey hands after handling PPE. Wash hands and face, shower if possible. 5.3 Contamination Control

5.3.1

Contamination Control Practices

Appropriate contamination control measures should be taken, when such measures do not interfere with necessary emergency medical attention. Factors influencing the need for contamination control measures: • injuries or other medical concerns (e.g., lacerations, smoke inhalation, embedded materials, burns, respiratory or cardiac distress, broken limbs, shock); • contamination of the individual (e.g., skin contamination, inhaled or ingested activity, contaminated clothing or hair); • contamination at the scene (e.g., spilled liquids, radioactive powder, and fallout); and • other dangers at the scene (e.g., fire, unstable structures, spilled chemicals, broken glass, dangerously high radiation levels, explosives). Contamination control will be a primary concern only if the contaminated person is lightly injured or uninjured and there are no other hazards at the scene. However, the presence of other significant hazards or serious injuries should take priority over contamination control. 5.3.2

Contamination Control of Exposed People

Contamination control should not interfere with caring for severely-injured people. Contamination control may include:

5.3 CONTAMINATION CONTROL

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• removing the person’s contaminated clothing; • wrapping the individual to control the spread of contamination; and • decontaminating the contaminated person. 5.3.3

Contamination Control Among Medical and EmergencyResponse Personnel

• medical and emergency-response personnel should not delay or deny treatment to contaminated persons; • following standard precautions (CDC, 1988) and wearing appropriate PPE will protect against contamination while working with a contaminated person (normally operating room precautions are adequate); and • changing PPE, using a step-off pad, and showering at the end of a shift is sufficient to prevent the spread of contamination into designated “clean” areas. 5.3.4

Radiologically-Controlled Areas (further defined in Sections 4.3.3 and 18)

Figure 5.1 illustrates important concepts of radiologically-controlled areas and their use. These apply to the sites of radionuclide releases (contamination incidents) and hospitals or other locations where contaminated patients are being examined, decontaminated or treated. • established boundaries should be at least waist high with clearly defined paths via step-off pads (e.g., rope, tape, or other device that can be controlled); • if possible locate a table with a barrier line separating clean and possibly contaminated surfaces at exit point for tools and equipment; • the markings on the floor at the designated exit points delineating the inner contaminated, outer contaminated, and secured areas should be low to permit crossing without tripping or losing balance. (To control contamination, persons should not cross boundaries anywhere except at designated points, the gates at the step-off pads.); • plastic sheeting (preferably textured to minimize slipping) should be used to cover the ground in the exit area; • “sticky mats” may be used as step-off pads to minimize transporting radioactive contamination from one area to the next; and • “hot” waste containers should be clearly marked “radioactive waste” and should be lined with plastic bags.

90 / 5. PERFORMING SURVEYS AND CONTROLLING CONTAMINATION

Fig. 5.1. Schematic drawing of radiologically-controlled areas and exit points at the scene of an incident.

5.3.4.1 General Guidelines for Operation of a Controlled Contamination Area • seriously injured patients and necessary medical personnel should use the contamination control corridor for evacuations to ambulances; • the inner and outer contaminated, and secured areas of the exit point should be used in the following manner: - the inner contaminated area and objects within are assumed to be contaminated at all times; - persons wishing to exit the inner contaminated area should do so only at an exit location such as that shown in Figure 5.1 unless their medical condition dictates otherwise; - the outer contaminated area is established for the removal of contamination control PPE, performing whole-body surveys, and necessary decontamination. This area should be surveyed and decontaminated frequently, if possible; - exit to the “clean” secured area should be restricted to those who are confirmed to be uncontaminated via wholebody survey and (if necessary) decontamination; - step-off pads (preferably consisting of “sticky mats” when available) should be used when stepping into the next area through the control points; - step-off pads should be surveyed periodically for contamination and should be replaced, decontaminated or renewed if found to be contaminated; and - contaminated clothing, tools, and other objects should be placed in radioactive-waste containers, in individual bags if possible.

5.3 CONTAMINATION CONTROL

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5.3.4.2 Leaving a Controlled Area 1. 2. 3. 4. 5.

6. 7. 8. 9. 10.

enter “hot” side of exit point; log name of responder entering the exit point; survey outer gloves or hands for contamination; survey coveralls or outer clothing for contamination; if contaminated, remove coveralls or outer clothing and place in radioactive-waste container or plastic bag (consistent with common practice, surveys prior to removing outer clothing may not be required as it doubles the survey time; in these cases, PPE should be removed at the step-off pad); step to “cold” boundary of exit point; remove shoe covers while stepping over boundary to “cold” side of exit point; remove gloves inside out and place into radioactive-waste container or plastic bag; survey whole body, concentrating on hands, feet, face, knees, elbows, and seat of pants; and survey exit point and step-off pad(s) periodically and decontaminate as necessary.

5.3.4.3 Transportation of Injured and Contaminated Individuals. Persons who require immediate medical care, prior to external decontamination onsite, can be transported to a hospital or other medical facility in conventional ambulances. The removal of contaminated clothing and initial skin decontamination at the incident site will eliminate most of the readily transferable contamination. In transportation accidents where no onsite decontamination is possible, placing the individual in a sheet or blanket and covering the litter and surrounding floor with plastic sheeting should prevent serious contamination of the ambulance or its attendants. Only in unusual cases would the driver or other personnel be at risk of being seriously contaminated. Nevertheless, the ambulance and the attending emergency medical responders should be monitored before they leave the hospital area if possible. If it is necessary to transport patients beyond the local area, by airplane or helicopter, similar precautions should be taken to minimize contamination of the vehicle. Ambulance and treatment area contamination control: 1. 2.

Wrap patient in blankets to contain contamination and reduce contamination of facilities. Establish dedicated routes for transporting contaminated patients.

92 / 5. PERFORMING SURVEYS AND CONTROLLING CONTAMINATION 3. 4.

5. 6.

7. 5.3.5 1.

2.

3. 4. 5. 6. 7.

8.

Establish dedicated areas for decontamination and contaminated patient care. Line dedicated routes and rooms with plastic to reduce contamination of fixed surfaces. Avoid use of slick plastic on floors and walking surfaces. Textured plastic or other nonslippery surface is preferred. Do not use contaminated vehicles or equipment for noncontaminated patients unless necessary. Use a “hot” stretcher to transport the individual to the boundary of the radiologically-controlled area, transferring to a “clean” stretcher for transport to the hospital. Wrap the patient in a blanket or sheet prior to placing on “clean” stretcher. Decontamination of Equipment Smooth surfaces (glass, plastic, metal) can be decontaminated using any of these methods: a. wipe with cloths dampened with water or commercial cleaning products; b. “tape press” using the sticky side of tape (duct tape, masking tape, etc.); and c. immersion in an ultrasonic cleaning sink. Contamination that cannot be removed can be considered fixed; if radiation dose rate is low, essential equipment may continue to be used. Porous or fibrous surfaces cannot be decontaminated by washing or wiping. Some items may be soaked in a cleaning solution or placed in an ultrasonic sink. Soft items (e.g., wood, plastic, lead) may be shaved with a knife to remove contamination. Contaminated sections of fabric or paper can be cut out and the remainder used. If contamination is fixed in equipment, and the equipment must be used, cover the contaminated area and continue using the equipment as necessary. Large areas (such as ambulance interiors or floors) may be decontaminated by wiping with a sponge or rags soaked in soapy water, detergent, or other cleaning solutions.

Part B: Onsite and Prehospital Actions 6. Stage 1: Medical Assessment (onsite triage area)

Objectives • identify individuals with life-threatening problems, stabilize and consider admitting for emergency care; • identify and treat individuals with nonlife-threatening injuries; • identify individuals exposed to radionuclides and external radiation; • identify those who may be externally and/or internally contaminated; • identify those who show evidence of psychological distress issues discussed in Section 9.1.2; and • document the radionuclide contaminating incident. 93

94 / 6. STAGE 1: MEDICAL ASSESSMENT (ONSITE TRIAGE AREA) See Figure 3.1 for flow of persons through all nine stages in the management of radionuclide contamination. The management of persons contaminated with radionuclides will usually occur at two locations, at the site of the incident and at a hospital or other emergency facility. In this Report, the actions taken by qualified experts, medical and radiation-safety professionals are considered to occur in nine stages. The flow of persons through these nine stages is illustrated in the chart given in Figure 3.1. The following describes actions to be taken in the first stage, Medical Assessment, at the site of the contamination incident. 6.1 Initial Actions of Medical and Radiation Safety Personnel • Remove individuals from the contaminated areas to a triage area within the decontamination area (Figure 4.1) • Remove contaminated clothing and replace with clean clothing or wrap in blanket. • Survey individual for surface-contamination levels and document. • Obtain nasal swabs and survey for evidence of an inhalation intake (see Section 10.3.1.1 for information on how to collect samples and to interpret the results and Section 6.6. for actions to be taken when swabs are positive). • Treat injuries and other medical concerns such as psychological distress issues discussed in Section 9.1.2. • Cover contaminated wounds with sterile dressings. • Alert hospital and call for ambulance service as necessary. • Take individual to the hospital if injuries require immediate medical or surgical care not available onsite or if further medical or dosimetric evaluation and treatment is required. • Take precautions to prevent spread of contamination during transport and movement of the patient. Have transport vehicles, attendants and equipment checked for residual radionuclide contamination before release from control area. • Advise family on the extent of injuries and exposure (Appendix E). • Obtain complete history of incident, especially as it relates to the activities of the individual. Where was he/she? What was he/she doing? Exit path? Symptoms (Section 6.8 and Appendix A). 6.2 Potential Life-Threatening Problems The highest priority is to provide emergency medical care immediately for serious injuries, prevent further injuries, and preserve

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vital functions. This has priority over assessing contamination and decontaminating (Bushberg et al., 2007). Minor injuries can wait until after initial radiation survey has been completed. • Medical personnel should consider both medical and radiological issues so that critical medical concerns (e.g., injuries that threaten life, limb or vision) are addressed first. • Contaminated or irradiated patients do not pose a health risk to medical responders, except in extremely unusual situations where the patient is very heavily contaminated with radionuclides that may become airborne, or may have fragments of highly-radioactive materials embedded in the skin or clothing. • Unless knowledgeable about the teratological effects of radiation and the dose to the fetus, medical personnel should not advise pregnant women on the medical effects of their suspected radiation exposure until they have had the opportunity to consult with trained radiation-safety personnel, medical physicists, or other professionals who are competent to perform a fetal radiation dose estimate and to interpret the results. • The highest priority usually is to provide immediate emergency medical care to individuals who have been seriously injured in a radionuclide contamination incident. However, should there be an ongoing hazardous situation, such as a fire, evacuation may be the highest priority. When a life or death surgical emergency exists, the patient must receive immediate lifesaving first aid and transportation to a hospital regardless of contamination, except as noted below. A hospital emergency room or surgical suite can always be decontaminated after its use. Skin or wound contamination is almost never immediately life threatening. • In rare situations, it may be necessary to perform emergency amputation or extensive surgical debridement at the site of the incident, in the nearest first aid station, at a decontamination facility, or medical-response base established onsite (Figure 4.1). For example, in the case of an explosion, an individual could suffer a mangled extremity contaminated with embedded gamma-emitting foreign bodies such that the radiation level to the rest of the body and to first aid personnel is an overriding consideration [e.g., gamma levels of Gy h–1 (hundreds of rad h–1)]. High-level contamination of this magnitude provides the principal need and justification for use of mobile gamma shields to protect persons engaged

96 / 6. STAGE 1: MEDICAL ASSESSMENT (ONSITE TRIAGE AREA) in decontamination work or surgery in radiological emergency centers. However, since such incidents are expected to be extremely rare, it is not likely that mobile shielding sufficient for gamma rays >500 keV would be a stock item at most emergency centers; only those associated with facilities where work with radiation sources was routine. • Unless the exposed person has serious injuries or the contamination incident site is poorly equipped or staffed, there is usually no need to move the patient immediately to a hospital. Even with fairly serious injuries, it is better to stabilize the patient’s physical condition and remove all easily transferable contamination than to rush him/her to a poorly prepared hospital. 6.3 Identification of Individuals Exposed to Radiation and/or Radionuclides Preliminary assessment of radiation injuries and external and internal depositions of radionuclides onsite will necessarily be qualitative. However, information and survey instrumentation data gathered onsite will be valuable resources for complete medical and dose assessments conducted at the hospital. Table 6.1 suggests methods useful in identifying exposed persons who may have been irradiated from external sources, those who may be externally contaminated and those who may have had intakes of radionuclides and are internally contaminated. Many assessment methods can be conducted onsite (field measurement). These should be given first consideration. However, if these are not adequate to make an initial assessment, then samples will need to be collected and sent to an appropriate laboratory for analysis. Section 5.1 provides guidance on performing surveys for evidence of external contamination and internal deposition. 6.4 Assessment of External Irradiation Certain radionuclide contamination scenarios can result in exposures to penetrating radiation from external sources (Section 17). When this is a possibility, it is important to determine whether exposures occurred and to evaluate possible radiation doses. If the incident occurs in a facility where personal dosimeters are in use, the evaluation is straight forward. However, when personal dosimeters or other radiation measurement devices are not present, the evaluation must rely on clinical and biological indicators. Table 6.2 is a concise description of the clinical characteristics of different acute radiation syndromes from high-dose, whole-body radiation exposures delivered at high dose rates.

TABLE 6.1—Preliminary assessment of radiation exposures, doses and contamination. Source

External

Assessment Method

Assessment Location

Measure radiation levels near individuals and note their stay-time

Field measurements; radiation survey or dosimeter

Skin or clothing contamination

Measure radiation dose rate from contamination with survey instrument

Field measurement

Skin

Skin dose calculations based on dose coefficients (see electron constants in Table 7.1 and Section 21.3 for further information)

Inhalation (nasal and/or oral)

Monitor facial area for radioactivity; collect nasal swabs

Field measurement Collect in field and survey for preliminary evidence of inhalation intake; analyze in laboratory for further evaluation (Section 10.3.1.1)

Air samples

Field; for quick assessment, hold probe over sample surface. Count in field and/or laboratory

Obtain count rate from lungs (gamma only)

Field; for quick assessment, hold probe on chest or back to monitor for activity

Urine, fecal samplesa

Collect samples, analyze in laboratory

samplesa

Collect samples, analyze in laboratory

Ingestion

Urine, fecal

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Radioactive source

6.4 ASSESSMENT OF EXTERNAL IRRADIATION

Internal

Pathway

Source

Pathway

aIn

Assessment Location

Dermal absorption

Urine, fecal samplesa

Collect samples, analyze in laboratory

Absorption via a wound

Obtain count rate from wound

Count in field or in medical facility/laboratory

Urine, fecal samplesa

Collect samples, analyze in laboratory

Obtain count rate at site of injection (or wound)

Survey in field or laboratory

Urine, fecal samplesa

Collect samples, analyze in laboratory

Calculate dose using lymphocyte depletion or cytogenetic assays after 24 h

Over 1 to 2 d, collect and process samples in laboratory and perform dose assessment

Utilize software such as the AFRRI Biodosimetry Assessment Tool (AFRRI, 2007)

Assessment may be performed in field or in hospital

Injection (e.g., contaminated shrapnel, and debris)

Internal, external

Assessment Method

Whole-body radiation exposure

vivo bioassay samples, urine and/or fecal, should be obtained in all cases of contamination or exposure to an unsealed source of radioactive material to assess radionuclide uptake. However, in vivo bioassay samples collected shortly following an intake (e.g., within 2 to 4 h for urine and within a couple of days for feces) are not necessarily representative of a systemic uptake. Further, collection of samples onsite that are free of contamination is often difficult. Therefore, it is usually prudent to move potentially-contaminated persons to a noncontaminated control area (e.g., medical facility, before collecting samples). However, early urine and blood samples can in some circumstances be useful in gaining information about the initial intake and later for comparison with “after treatment” samples (Section 10.3).

98 / 6. STAGE 1: MEDICAL ASSESSMENT (ONSITE TRIAGE AREA)

TABLE 6.1—(continued)

TABLE 6.2—Acute radiation syndromes (CDC, 2005). Latent Stage

Manifest Illness Stage

Recovery

Hematopoietic (bone marrow)

>0.7 Gy (>70 rad) [mild symptoms may occur as low as 0.3 Gy (30 rad)]

• Symptoms are anorexia, nausea and vomiting • Onset occurs 1 h to 2 d after exposure • Stage lasts for minutes to days

• Stem cells in bone marrow are dying, although patient may appear and feel well • Stage lasts 1 to 6 weeks

• Symptoms are anorexia, fever, and malaise • Drop in all blood cell counts occurs for several weeks • Primary cause of death is infection and hemorrhage • Survival decreases with increasing dose • Most deaths occur within a few months after exposure

• In most cases, bone-marrow cells will begin to repopulate the marrow • There should be full recovery for a large percentage of individuals from a few weeks up to 2 y after exposure • Death may occur in some individuals at 1.2 Gy (120 rad). • LD50/60b is ~2.5 to 5 Gy (250 to 500 rad)

Gastrointestinal

>10 Gy (>1,000 rad) [some symptoms may occur as low as 6 Gy (600 rad)]

• Symptoms are anorexia, severe nausea, vomiting, cramps, and diarrhea • Onset occurs within a few hours after exposure • Stage lasts ~2 d

• Stem cells in bone marrow and cells lining GI tract are dying, although patient may appear and feel well • Stage lasts <1 week

• Symptoms are malaise, anorexia, severe diarrhea, fever, dehydration, and electrolyte imbalance • Death is due to infection, dehydration, and electrolyte imbalance • Death occurs within 2 weeks of exposure

• LD100c is ~10 Gy (1,000 rad)

/ 99

Prodromal Stage

6.4 ASSESSMENT OF EXTERNAL IRRADIATION

Dosea

Syndrome

Syndrome Cardiovascular / central nervous system

Dosea

Prodromal Stage

>50 Gy (5,000 rad) [some symptoms may occur as low as 20 Gy (2,000 rad)]

• Symptoms are extreme nervousness and confusion; severe nausea, vomiting, and watery diarrhea; loss of consciousness; and burning sensations of the skin • Onset occurs within minutes of exposure • Stage lasts for minutes to hours

Latent Stage • Patient may return to partial functionality • Stage may last for hours but often is less

Manifest Illness Stage • Symptoms are return of watery diarrhea, convulsions, and coma • Onset occurs 5 to 6 h after exposure • Death occurs within 3 d of exposure

Recovery • No recovery is expected

aThe absorbed doses quoted here are “gamma equivalent” values. Neutrons or protons generally produce the same effects as gamma, beta, or x rays but at lower doses. If the patient has been exposed to neutrons or protons, consult radiation experts on how to interpret the dose. bThe LD 50/60 is the dose necessary to kill 50 % of the exposed population in 60 d. c The LD100 is the dose necessary to kill 100 % of the exposed population.

100 / 6. STAGE 1: MEDICAL ASSESSMENT (ONSITE TRIAGE AREA)

TABLE 6.2—(continued)

6.5 CONTAMINATION SCREENING OF INDIVIDUALS

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Patients who develop nausea and vomiting in the first 6 to 12 h after a radiation incident should be hospitalized. Since nausea and vomiting rarely occur as an emotional reaction to a radiation incident, they should be considered indicative of a serious exposure to penetrating radiation until proved otherwise. A complete blood cell count [CBC (total and differential)] should be performed promptly and then every 6 to 8 h as indicated. If the counts reveal a rapid fall or a low value in absolute lymphocyte count within 48 h, radiation injury is strongly suggested. Other laboratory diagnostic techniques, such as cytogenetic chromosome analysis may be useful. Except in cases of extremely-high exposure, radiation injury typically has a latency period of days to weeks (see Table 6.2 for signs and symptoms resulting from various levels of whole-body irradiation). Skin burns, vomiting, weakness, and other apparent radiation effects that appear after a short time may be caused by other factors. 6.5 Contamination Screening of Individuals 6.5.1

External Contamination

Depending upon the nature of the contaminating incident, it may be that not all persons onsite will be exposed and contaminated. Therefore, the first step is to identify the contaminated individuals. This screening should occur onsite using methods given in Table 6.1. All individuals in the area of radionuclide release incident should be surveyed with appropriate radiation detection instruments identified in Tables 5.1 and 5.2. These should be total-body surveys conducted in an uncontaminated area if possible. The results should be recorded on a body diagram (Appendix A). Individuals found to have external contamination should be admitted to a controlled area for assessment of their contamination (Section 7) and for decontamination (Section 8). 6.5.2

Internal Contamination

It should be assumed that those who have been determined to be externally contaminated also have had an intake of radionuclides and are internally contaminated. Internal depositions of radionuclides may result from inhalation, ingestion, skin absorption, contaminated wounds, and/or embedded radioactive shrapnel and other contaminated debris. An assessment of the incident and the triage findings should be well documented by radiation-safety and medical personnel (Appendix A). Most contamination incidents take place at facilities with radioactive-materials licenses (nuclear power plants, research

102 / 6. STAGE 1: MEDICAL ASSESSMENT (ONSITE TRIAGE AREA) universities, etc.), which are likely to have the capability to perform all of the following actions. In the event of a large-scale contamination incident (e.g., a terrorist attack), contamination may be spread to public areas with limited resources for rapid and appropriate response to radiological emergencies. Accordingly, it may not be reasonable to assume that all of the following actions will be accomplished rapidly by trained radiation-safety professionals. Once the radiological character of an incident is recognized, emergency responders should immediately request support from radiationsafety professionals, and should take those actions that they can while waiting for their arrival. When possible, direct-body measurements appropriate to the radionuclide of exposure and the mode of potential intake can be made onsite using methods identified in Table 6.1. These should provide useful information about the likely magnitude of intake. Initial indications of an intake may include external contamination in the vicinity of the mouth or nose, air sample results, a wound incurred while working with radioactive material, skin breaks or blood smears showing radioactive contamination, and oral or nasal swabs (or nasal blows) showing contamination. Such initial indicators should usually be considered qualitative rather than quantitative for the purposes of intake and dose assessment. However they may be adequate for making an initial determination whether dose-reduction therapy (i.e., treatment with medical countermeasures) is indicated. The possibility of false-negative results, such as from oral swabs because of rapid clearance from the mouth, should not be overlooked. More accurate measurement of internal contamination is dependent upon the collection of appropriate biological samples (Section 10). 6.5.2.1 Inhalation Intakes. Nasal-swab samples can be useful in providing qualitative information about an inhalation intake. However, collecting such samples may be limited in situations where there are many contaminated individuals. When collecting nasal swabs is practical, the swab samples should be obtained on anyone suspected of inhaling radioactive material, as soon as the person’s condition permits, and prior to showering or washing the face. The samples must be collected early to obtain an accurate result, before an individual puts a finger in, or blows, his/her nose (see Section 10.3.1.1 for guidance in collecting and interpreting nasal swabs). Nasal swabs should be surveyed with a GM or sodium iodide detector for beta/gamma emitters and a zinc-sulfide detector for alpha emitters. Positive results are indicative of possible intakes and should be followed by further sampling and laboratory analysis.

6.5 CONTAMINATION SCREENING OF INDIVIDUALS

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Individuals with intakes of radionuclides should be further assessed by radiation-safety professionals at an appropriate facility after external decontamination. 6.5.2.2 Intakes Through Skin and Ingestion. Wounds in areas of skin contamination are strong indicators of possible radionuclide intakes. Contamination of skin surfaces suggest possible intakes through absorption, but only in cases of very heavy contamination would absorption result in significant internal contamination, even if the radionuclide is in a soluble form. Injuries with contaminated debris and shrapnel are clear evidence of internal contamination. Discovery of contaminated food and water should be taken as evidence of possible ingestion intakes. Contamination of the mouth and other oral surfaces suggest possible intakes by ingestion, not necessarily in contaminated food, but by touching the face and mouth with contaminated hands. • Survey the individuals’ mouths and nostrils with a GM, sodium iodide (for gamma radiation), or zinc-sulfide (for alpha radiation) detector for evidence of inhalation or ingestion (Section 5.1.1). • Survey wounds with a GM, sodium iodide (for gamma radiation), or zinc-sulfide (for alpha radiation) detector for evidence of contamination or shrapnel. • Attempt to determine the time of intake as accurately as possible. • Emergency medical personnel (with radiation-safety personnel guidance if needed) should attempt to collect samples for bioassay as tabulated in Table 6.1 for internal dose assessment. Such onsite samples should rarely include urine and fecal samples, as noted below and in Table 6.1. An accurate determination of internal contamination may take several hours to obtain and requires urine and stool bioassays or other procedures described in Section 10. 6.5.2.3 Collection of Excreta. Although it is important to collect samples for bioassay soon after a contamination incident, urine samples taken within 4 h and fecal samples within a couple of days are of little value because they may not be representative of the systemic uptake. In addition, samples taken in or near contaminated areas have a high probability of being contaminated.

104 / 6. STAGE 1: MEDICAL ASSESSMENT (ONSITE TRIAGE AREA) 6.6 Onsite Treatment for Internal Contamination Only rarely is it necessary to initiate decorporation therapy at the site of a serious radionuclide contamination incident. Treatment should be limited to simple actions such as oral, nasopharyngeal, and wound irrigation to remove radionuclides that might be ingested, taken further into the respiratory tract, or absorbed into the tissues surrounding wounds. Prompt administration of medical countermeasures or isotopic dilution can appreciably decrease the uptake of radionuclides into stable metabolic pools such as thyroid and bone, from which it is not possible to readily mobilize radionuclides. In order to be effective, these agents must be given as soon as possible after the intake (Section 12.2). For example, if intakes of plutonium or other actinides are suspected (e.g., from surveys of the facial areas or by positive nasal swabs), DTPA should be administered immediately (Breitenstein, 2003; IAEA, 1978; NCRP, 1980; Norwood, 1975) (Section 12.3.3). Also, if intakes of radioiodine are suspected, administration of the blocking agent, KI should be considered. The physician (or paramedic) at the site of the contamination incident should therefore administer the appropriate agent as soon as possible if it has been determined that intake levels were high. If the appropriate treatment agent is not available onsite, the physician should contact the hospital and suggest its administration as soon as the patient arrives. Specific recommendations for drug treatment are described in Section 12. Table 12.2 provides a rapid means of finding the appropriate agent to use for a particular exposure. 6.7 Priorities in Processing Exposed Persons The objective of Stage 1 is the identification and assessment of exposed individuals. This should lead to the characterization of each as to the degree of their injury and contamination. Table 6.3 suggests priorities to be considered in processing individuals through succeeding management stages. 6.8 Documenting a Radionuclide Contamination Incident Onsite physicians and/or other medical personnel should forward information on the radionuclide incident to the emergency department of their referral hospital and remain in contact with the hospital staff until treatment decisions have been made. Medical and radiation-safety personnel should carefully document the incident including their actions and observations. An example form for documenting radiological exposures is in Appendix A.

TABLE 6.3—Priorities in processing exposed persons. Degree of Injury and Contamination Level

Actions

High external radiation doses

Transport to hospital before decontamination and treat for acute radiation syndrome.

High levels of internal contamination (thousands of disintegrations per minute in nasal swabs from inhaled radionuclides, or thousands of counts per minute from contaminated wounds)

Intervene, if possible, to enhance the total-body natural elimination rate of the compound, or block the uptake of the radionuclide in sites where high uptake may occur (e.g., radioiodine in the thyroid and plutonium in bone). As noted above, if intakes of plutonium or other actinides are confirmed or suspected, immediate treatment with DTPA should be started if possible (Section 12) and if intakes of radioiodine are indicated, administrations of KI should be started. If immediate intervention is contra-indicated, decontaminate and admit to emergency facility for internal contamination and dose assessment (see Section 8 for decontamination guidance).

Serious injuries (not life threatening, but requiring rapid medical attention)

Decontaminate and admit to emergency department.

Evidence of contaminated wounds

Gently flush with sterile water or saline (see Section 9.4.1 for further information).

High levels of skin contamination (Tables 3.8 and 7.2) [>10,000 Bq cm–2 (>600,000 dpm cm–2) beta/gamma surface body contamination (IAEA, 2005b), or spot contamination >3,700 Bq (2.2 × 105 dpm) (NCRP, 2005a)]

Decontamination required and admit to emergency department for treatment of skin radiation burns (Section 8). Evaluate for internal contamination.

/ 105

Manage airway, breathing and circulation and transport to hospital emergency department before decontamination (see Section 5.3.4 for transportation of contaminated patients). 6.8 DOCUMENTING A CONTAMINATION INCIDENT

Life-threatening injuries and medical conditions

Degree of Injury and Contamination Level

Actions

Moderate injuries (requiring medical attention)

Decontaminate and admit to emergency department.

Moderate levels of internal contamination (hundreds of disintegrations per minute in nasal or oral swabs, or hundreds of disintegrations per minute from contaminated wounds)

Decontaminate and admit to emergency department for internal contamination and dose assessment.

Moderate levels of surface body contamination [>1,000 Bq cm–2 (>60,000 dpm cm–2) beta/gamma]

Decontamination advisable and evaluate for internal contamination (Table 7.2).

Mild injuries with low levels of surface body contamination [>100 Bq cm–2 (>6,000 dpm cm–2) beta/gamma]

Decontaminate to at least two times background and discharge if no evidence of internal contamination (Table 7.2).

Low levels of surface body contamination [>100 Bq cm–2 (>6,000 dpm cm–2) beta/gamma] but uninjured

Decontaminate to at least two times background and discharge if no evidence of internal contamination (Table 7.2).

Individuals with only surface body contamination [<100 Bq cm–2 (6,000 dpm cm–2) beta/gamma (IAEA, 2005b) or less than ~170 Bq cm–2 (<10,000 dpm cm–2) (NCRP, 2005a)]

Can be released.

106 / 6. STAGE 1: MEDICAL ASSESSMENT (ONSITE TRIAGE AREA)

TABLE 6.3—(continued)

7. Stage 2: External Contamination Assessment (onsite triage area)

Objectives • determine location, quantity and identity of external contamination; • identify hot particles, shrapnel, and other contaminated debris; • evaluate possible radiation-induced dermal and subdermal injury; • assess possible internal intakes of radionuclides deposited on skin; and • assess need for treatment and decontamination. See Figure 3.1 for flow of exposed persons through all nine stages in the management of radionuclide contamination. 7.1 External Contamination Assessment Procedures Characterization of external contamination is necessary to assess the risk of radiation injury to the skin and underlaying tissues, and 107

108 / 7. STAGE 2: EXTERNAL CONTAMINATION ASSESSMENT to determine the potential for internal contamination occurring as a result of absorption of radionuclides through the skin or from wounds or from intake by inadvertent ingestion and inhalation of radionuclides deposited on the skin and clothing. Characterization is also necessary for making decisions about treatment and decontamination. While a complete qualitative and quantitative characterization is ideal, the timely removal of external contamination usually takes precedence over any in-place characterization beyond a simple contamination survey. Radionuclide identification or detailed characterization by energy or material form may be more effectively accomplished using contamination material from a smear sample or clothing sample. Further information is available in several documents (FEMA, 2002; IAEA, 2006; NCRP, 2005a) and the REMM website (DHHS, 2009). The level of external contamination will determine the procedures and instrumentation needed for an accurate assessment of the nature of the contamination, its location on the body and the radiation dose rates to the skin and internal tissues. This assessment should take place away from the immediate contaminated incident site in an outer contaminated area or in a triage area (Figure 4.1). The specific objectives are: • locate areas of contamination on the body including body orifices (marking them may prove helpful for decontamination and subsequent monitoring); • locate contaminated wounds; • identify the nature of the contaminating sources, such as hot particles, shrapnel, and other debris; • determine whether skin contamination is “loose” and susceptible to being removed by washing or changing clothes or “fixed” in place (FEMA, 2002); • identify the contaminating radionuclides; • assess quantity of radionuclides and location, whether spread over the body or localized; • evaluate potential for skin injury; • confirm identification of those individuals with possible internal contamination; and • identify individuals for treatment and decontamination consideration. The following steps are recommended to achieve these objectives: 1.

Perform a complete body survey (see Section 19.4 for information on survey instrumentation).

7.2 DOSE ASSESSMENT

2. 3.

4. 5.

6.

7.

8. 9.

10. 11.

/ 109

Remove contaminated clothing and replace with clean clothing or wrap in sheet/blanket as needed. Survey individual completely with a radiation survey meter for surface contamination, identifying the locations of the contamination on the body and the levels. Assume the contamination is “loose” and susceptible to being spread around until proven to be “fixed” to the skin. Examine for contaminated shrapnel and other debris. Identify the contaminating radionuclides with a portable spectroscopic survey meter. A smear sample or clothing sample may facilitate radionuclide identification and allow more timely personal decontamination. Section 3.2.2 describes methodology for characterizing contamination using a GM detector in the absence of spectrometry equipment. Survey hair and body orifices such as eyes, ears and skin folds, for radionuclide contamination including hot particles using a detector appropriate for the contamination of concern (e.g., GM detector for beta/gamma radiation, sodium iodide detector for gamma radiation, zinc-sulfide detector for alpha radiation). Examine the individual for wounds and abrasions that could be routes for intradermal and internal radionuclide depositions. Survey wounds and abrasions with a detector appropriate for the radiation of concern. Survey the individual’s face with a detector appropriate for the radiation of concern as evidence of ingestion or inhalation intakes. Take nasal swabs and survey for evidence of inhalation intake. Document external contamination (see Table A.3.1). 7.2 Dose Assessment

Irradiation of the skin can occur from radionuclides deposited on the skin. It can also occur from sources located away from the body. However, estimates of the latter require information about radionuclides in the environment in which the individuals were at the time of exposure that is rarely available. Therefore, assessment of skin doses resulting from exposure to radionuclides focuses on radionuclides deposited upon the skin. These skin doses can be estimated using body survey data and the electron constants from Table 7.1. Section 21.3 gives guidance in interpreting external monitoring data.

110 / 7. STAGE 2: EXTERNAL CONTAMINATION ASSESSMENT TABLE 7.1—Electron constants (skin dose rate at depth of 70 μm from source on skin surface) for selected radionuclides. Electron Constantc Half–Life

Decay Modeb

P

14.3 d

60Co 90

Radionuclidea

[Gy s–1 (Bq cm–2)–1]

[rad h–1 (μCi cm–2)–1]

β–

6.55 × 10–10

8.72 × 100

5.27 y

β–

3.09 × 10–10

4.12 × 100

28.8 y

β–

4.98 × 10–10

6.63 × 100

90Y

64.1 h

β–

6.65 × 10–10

8.86 × 100

106Ru

374 d

β–

0d

0d

*106Rh

29.8 s

β–

6.98 × 10–10

9.30 × 100

103Pd

17 d

EC

2.87 × 10–15

3.82 × 10–5

131I

8.0 d

β–

4.82 × 10–10

6.42 × 100

137

Cs

30.2 y

β–

4.73 × 10–10

6.30 × 100

*137mBa

2.55 month

IT

6.88 × 10–11

9.16 × 10–1

144Ce

285 d

β–

2.74 × 10–10

3.65 × 100

192

73.8 d

β– EC

5.46 × 10–10

7.27 × 100

210Po

138 d

α

8.53 × 10–17

1.14 × 10–6

227Ac

21.8 y

β– α

1.79 × 10–13

2.38 × 10–3

U

4.47 × 109 y

α SF

2.73 × 10–13

3.63 × 10–3

*226Ra

1,600 y

α

1.42 × 10–11

1.89 × 10–1

238Pu

87.7 y

α SF

2.49 × 10–13

3.32 × 10–3

239

24.1 y

α

1.29 × 10–13

1.72 × 10–3

241Am

432 y

α

7.60 × 10–13

1.01 × 10–2

244Cm

18.1 y

α SF

7.78 × 10–14

1.04 × 10–3

252

2.65 y

α SF

1.32 × 10–10

1.76 × 100

32

Sr

238

Ir

Pu

Cf

aNames preceded by an asterisk are radioactive progeny that may be present in significant quantities. b EC = electron capture IT = isomeric transition SF = spontaneous fission c The values tabulated for radionuclides undergoing spontaneous fission include the contribution from delayed beta emissions. dBeta emission of insufficient energy to penetrate to a depth of 70 μm.

7.4 TREATMENT GUIDANCE

/ 111

7.3 Screening Guidance Operational intervention levels for skin contamination are given in Table 7.2 (IAEA, 2005b). These default values are suggested for use where “surface contamination derived limits are not specified by the national competent authority.” For individuals with skin levels <10 Bq (0.27 nCi) (600 dpm) alpha radiation per square centimeter or 100 Bq (2.7 nCi) (6,000 dpm) beta/gamma radiation per square centimeter, no action is required and they may be released. Decontamination is advisable when skin levels are >100 Bq cm–2 (2.7 nCi) alpha or 1,000 Bq cm–2 (27 nCi) beta/gamma and required at levels 10 times these values. NCRP (2005a) and FEMA (2002) guidance is to give decontamination priority to individuals with spot contamination >37,000 Bq (1 μCi) to avoid skin injury. The presumed spot size is 0.2 cm2. 7.4 Treatment Guidance The presence of dermal injuries such as skin burns, abrasions and wounds should be followed with further assessment and possible treatment. Individuals showing evidence of receiving skin doses >2 to 4 Gy (200 to 400 rad) averaged >1 cm2 at 70 μm depth have about a 10 % chance of clinically detectable effects (NCRP, 1999) and should be followed for possible dermal injuries (Section 16.7.1.6). Decontamination should precede treatment, unless precluded by the seriousness of the injuries.

Alpha {Bq cm–2 (nCi cm–2) [dpm cm–2]}

Beta/Gamma {Bq cm–2 (nCi cm–2) [dpm cm–2]}

Beta/Gamma [(low background area)a μSv h–1 (μrem h–1)]

<10 (<0.27) [<600]

<100 (<2.7) [<6,000]

Not detectable

None • allow release

>10 (>0.27) [>600]

>100 (>2.7) [6,000]

Not detectable

Intervention optional • decontaminate or advise to shower and wash clothing • no significant health risk • slow release

>100 (>2.7) [>6,000]

>1,000 (>27) [>60,000]

0.2 – 0.3 (20 – 30)

Intervention advisable • prevent inadvertent ingestion and inhalation, limit spread of contamination and decontaminate

>1,000 (>27) [>60,000]

>10,000 (>270) [>600,000]

2–3 (200 – 300)

Intervention required • prevent inadvertent ingestion and inhalation, limit spread of contamination and decontaminate

aAmbient

dose equivalent rate measured at 10 cm from skin surface.

Actions

112 / 7. STAGE 2: EXTERNAL CONTAMINATION ASSESSMENT

TABLE 7.2—Skin contamination intervention levels (adapted from IAEA, 2005b).

8. Stage 3: External Decontamination (onsite decontamination area)

Objectives • control external contamination to avoid internal intakes and contamination of personnel and facilities; • reduce radiation dose to skin and risk of dermal injuries; and • reduce amounts of radionuclides in wounds. See Figure 3.1 for flow of persons through all nine stages in the management of radionuclide contamination. 8.1 Decontamination of Persons Whenever possible, partial or complete external decontamination of injured patients should be performed at the site before they are sent to a hospital. All contaminated clothing should be removed. This will generally remove 80 to 90 % of the external contamination (Goans, 2004). Uninjured persons can frequently 113

114 / 8. STAGE 3: EXTERNAL DECONTAMINATION decontaminate themselves but they must be given suitable instructions and be carefully monitored by an individual experienced with decontamination techniques and the use of radiation survey instruments. If only localized areas of contamination, such as the hands or face, are involved, these should be cleaned by washing the area with detergent and water. In cases of more generalized contamination, the person is instructed to shower or, if not ambulatory, the person should be thoroughly washed with soap or detergent. Often an initial shower can be given near the incident site and the patient then moved to an emergency medical and decontamination area where more elaborate skin decontamination techniques can be used. In most cases, radiation levels can be reduced sufficiently so that patients can be managed with but a few precautions at the hospital. In cases of very cold weather, washing of skin or showering will need to be performed at an indoor facility. 8.1.1

Decontamination Objectives

The goals of skin decontamination are to “decrease the risk of acute dermal injury, lower the risk of internal contamination and reduce the potential of contaminating medical and other personnel and the environment” (DHHS, 2009). It is also important to reduce the risk of skin cancer. However, the mortality rate for ionizing radiation-induced skin cancer is only ~1 to 3 %. Melanoma, the malignant skin cancer having the highest mortality rate does not appear to be induced by ionizing radiation (Mettler and Upton, 1995). While the ideal objective would be to reduce external contamination to background levels, in practice this is difficult to achieve except in cases of very minimal contamination. Decontamination is considered successful when the quantities of radionuclides on the external body are reduced to recommended levels. It is recognized that large numbers of contaminated people could overwhelm decontamination capabilities. This is taken into consideration in the levels recommended for decontamination. Therefore, a practical goal of whole-body decontamination is to decrease the level of contamination to no more than two times background (DHHS, 2009). However, in the event of large numbers of contaminated people, decontamination capabilities may become exhausted and the goal of two times background becomes impractical. In these cases, another set of decontamination criteria is recommended. For deterministic effects, FEMA-recommended guidance is 3.7 kBq (0.1 μCi) or 222,000 dpm for fixed contamination on a spot of skin (0.2 cm2 or a circle 0.5 cm in diameter). If the contamination

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is mixed (loose and easily displaced plus contamination fixed in place on the skin), the guidance is 37 kBq (1 μCi) or 2.2 × 106 dpm for spot contamination (FEMA, 2002). For stochastic effects, FEMA guidance is 2.7 MBq (73 μCi) or 1.6 × 108 dpm for fixed contamination over the body regardless of distribution. For an adult with the contamination uniformly distributed over the body, this corresponds to 150 Bq cm–2 (0.004 μCi cm–2) or 9,000 dpm cm–2. If the contamination is mixed (loose and fixed) the guidance is a factor of 10 higher. Table 8.1 gives decontamination goals applicable when large numbers of persons are contaminated. The values in the column labeled “Spot” relate to the FEMA guidance for deterministic effects and the values under “Body Surface” relate to the FEMA guidance for stochastic effects. In the absence of published values for alpha radiation, a value one-tenth of the beta/gamma value is recommended. This is based on IAEA skin and clothing criteria for alpha emitters being one-tenth the beta/gamma criteria. (For alpha emitters, the concern is not the risk to skin but the potential for inhalation of dislodged particles.) These criteria “indicate the level of skin contamination which could represent a hazard from direct irradiation of the skin, from intake by inadvertent ingestion, or that could indicate that the person has already inhaled or ingested significant amounts of radioactive material” (IAEA, 2006). 8.1.2

Decontamination Procedures

The following steps are recommended: 1. 2. 3.

4.

5.

6. 7.

Take individual to an area where skin decontamination or showering can be done. Perform a radiological survey documenting area of contamination. Carefully remove clothing; this generally removes 80 to 90 % of contamination (AFRRI, 2003; Bushberg et al., 2007; Goans, 2004; Koenig et al., 2005). Repeat a radiological survey of the unclothed body; marking areas of contamination with permanent pen may be helpful. Document area(s) of contamination. If multiple areas are contaminated, decontaminate areas with open cuts or wounds first, mouth and nose next, and then contaminated skin beginning with the most-contaminated areas. Decontaminate the skin with lukewarm water and soap, do not scrub vigorously. Carefully remove contaminated shrapnel and other debris with forceps.

116 / 8. STAGE 3: EXTERNAL DECONTAMINATION TABLE 8.1—Decontamination guidance; applicable when large numbers of people are contaminated and the goal of less than two times background is impractical. Spota (0.2 cm2)

Body Surface

Alpha

<0.37 k Bq (0.01 μCi) 22,000 dpm

—

Beta/gamma

<3.7 kBq (0.1 μCi) 220,000 dpm

<170 Bq cm–2 (4.5 nCi cm–2) 10,000 dpm cm–2

Contamination

Reference

IAEA (2006)

FEMA (2002) NCRP (2005a)

aFor contamination fixed on skin, limit is factor of 10 greater for mixed loose and fixed.

8.

9.

Potentially-contaminated open wounds require special attention. NCRP Report No. 156, Development of Biokinetic Model for Radionuclide Contaminated Wounds and Procedures for Their Assessment, Dosimetry and Treatment (NCRP, 2006a) gives detailed information on decontaminating wounds. Flush with tepid sterile saline or water jet under mild pressure to remove as much contamination as possible without causing additional injury. If contamination with plutonium is suspected, consideration should be given to irrigating the wound with DTPA (Section 9.4.1). As with any other route of intake of plutonium or other actinides, consideration should be given to intravenous (IV), intramuscular (IM), or oral administration of DTPA (Section 12). Care should be taken to avoid over dosing with DTPA because the amount absorbed from wounds cannot be measured. If further decontamination of wounds is indicated, the patient should be admitted to hospital as soon as possible (see Section 9.4.1 for further information on decontaminating wounds). Decontaminate external body cavities: eyes, ears, nose and mouth with tepid saline or water. Large areas of contaminated skin may require a shower. Decontamination of intact skin usually requires only gentle scrubbing with soap and warm water. Use of hot water is contraindicated (not recommended) because of the subsequent vasodilation (increased blood flow to the skin) that could increase absorption of the radionuclide. If more aggressive decontamination is necessary, a mixture of half corn meal and half laundry detergent has been shown to be effective.

8.2 PERFORMING DECONTAMINATION PROCEDURES

10.

11. 12.

13. 14. 15.

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Hair can usually be decontaminated with soap and water. If this is inadequate, the hair should be clipped rather than shaved, to avoid disruption of the skin barrier (NCRP, 2006a). Resurvey to confirm decontamination efficacy (see Section 19.2.7.3). No more than two decontamination cycles are recommended by the REMM website (DHHS, 2009) although typical occupational worker practice is to continue as long as each cycle results in reducing contamination 50 % or until reddening of the skin occurs. Document steps and actions taken using forms similar to those found in Appendix A. Cover the decontaminated patient with clean blankets or clothing. Using clear gloves remove the decontaminated patient to a clean secured area using clean gloves on a stretcher or in a chair, as appropriate. 8.2 Guidance for Those Performing Decontamination Procedures

• Save contaminated clothing and other items from the site to analyze for particle sizes and isotopic information. • Personnel should wear surgical scrub suits, surgical caps and gowns, and rubber gloves (surgical, household or industrial depending upon duties). • Rubber or plastic shoe covers are desirable. Those performing the actual decontamination with water should wear plastic or rubber laboratory aprons. • Air conditioning and forced-air heating systems should be turned off when practicable unless equipped with highefficiency particulate air filters to remove airborne contamination. • The floors should be protected with a disposable covering to reduce “tracking” by keeping surfaces clean and to aid cleanup tasks. • All contaminated clothing should be placed carefully into plastic or paper bags to reduce secondary contamination of the area. • Splashing of solutions used in decontamination should be avoided. • Patients and other potentially-contaminated personnel may move to clean areas only after surveys show satisfactory decontamination.

118 / 8. STAGE 3: EXTERNAL DECONTAMINATION • All persons and property passing between contaminated and clean areas must be surveyed and regulated by monitoring teams. • Leaders of decontamination teams should be trained in radiation monitoring and decontamination techniques. • Fiberboard or steel drums with tight-fitting tops should be obtained for disposable contaminated materials (not those saved for analysis). Labels describing the contents should be affixed so that proper disposal can be carried out without reopening them. • Personal dosimeters [e.g., pocket chambers, thermoluminescent dosimeters (TLDs), or optically-stimulated luminescent dosimeters (OSLDs)] should be supplied to all personnel working in the decontamination area. • Personnel are required to be rotated after a dose of 50 mSv (5 rem), the dose limit for emergency workers (FEMA, 2008). 8.3 Decontamination Facilities Onsite decontamination facilities exist at some installations where radionuclides are handled routinely. Emergency decontamination facilities can be improvised in a locker room, or shower room at some facilities such as industrial plants, hospitals, athletic buildings, and health clubs. The one essential feature is a shower or a bathtub. A fully-functional decontamination facility would have: • convenient equipment to wash both ambulatory and injured persons; • portable or permanent shielding for use in treating persons with high-level beta/gamma activities; • provision for collecting contaminated waste water; and • floor plan that will permit convenient decontamination work with a minimal opportunity for cross contamination of areas in the building. Supplies for decontamination, which ideally should be stored ahead of time, are listed in Table 8.2. These decontamination facilities would be located in a noncontaminated secured area (Figure 4.1). Floors of facilities used for decontamination purposes should be covered with heavy paper or plastic, the area should be isolated, all nonessential items should be removed from the room, and the staff

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TABLE 8.2—Suggested supplies for contamination assessment and decontamination. • Coveralls or surgical scrub suits • Plastic aprons • Surgical caps • Plastic or rubber gloves • Plastic shoe covers • Sterile surgical gloves • Respirators (prefit for team personnel) • Personal dosimeters • Sterile suture sets with additional sterile scissors (two), forceps (four), scalpel (one), and hemostats (six) • Sterile irrigation sets • Sterile applicators and miscellaneous dressings • Clean long patient gowns or coveralls, socks • Large towels • Bandage scissors (two) • Large plastic or cloth bags for collection of contaminated clothing • Radiation tags, radiation tape for marking areas • Radiation area signs, “Do Not Enter” • Masking tape (two inches wide) • Labeled containers for collecting urine and fecal specimens • Blankets • Adhesive labels and tags for labeling tissue or contaminated material • Specimen bottles (with formalin if freezing facilities are not available) • Felt pens (black and red) • Note books, papers, pencils • Data forms [e.g., decontamination forms (Appendix A)] • Portable beta/gamma survey meters. Include low range (up to 500,000 cpm or 50 mR h–1) and high range instruments (up to 500 R h–1) • Portable alpha scintillation detector • One large roll white absorbent (blotter-type) paper or wrapping paper as used in stores (tear-off dispensers are available for convenient storage and use of paper rolls) • Plastic sheets • Specific decontamination supplies (primarily detergents) • Fiberboard barrels or large waste baskets for disposal of contaminated clothing as well as other contaminated items - disposable or other alternative clothing for individuals whose clothing is too contaminated to be worn • Resealable plastic bags (various sizes) • Equipment for blood samples • Catch basin for decontamination liquids • Hair clippers • Self-adhesive labels

120 / 8. STAGE 3: EXTERNAL DECONTAMINATION should be dressed in PPE before contaminated individuals are admitted (Section 5.2). PPE should be used and worn under the direction of trained radiation-safety personnel. Respirators should be easily available and worn if significant transferable contamination (especially alpha emitters) is anticipated. Unless the room has a separate air conditioning or ventilation system equipped with high-efficiency particulate air filters, the ventilation system should be shut off temporarily until the extent of transferable contamination has been determined and controlled. If control of ventilation is difficult, intake and exhaust vents can be covered with plastic. Where contamination levels may be high [MBq (μCi) levels] it is desirable that all wash water empty into a special process drain for radioactive contamination disposal or into holding tanks. However, inability to divert wash water from the domestic sewer system should not delay or retard the decontamination effort in an emergency. 8.4 Saving Contaminated Materials Sponges, applicators and instruments that have been used to probe or cleanse any contaminated wound should be kept in separate containers identified as to source and sequence. Each excised tissue specimen should be monitored for radioactive contamination, put in a clean separate container, and frozen, if possible. If freezing is not available, specimens may be put into fixatives normally used for surgical pathology specimens. In most cases, the amount of readily removable radioactive contamination on the skin or in the wound site, especially when initial decontamination has been done onsite, is too small to justify holding the wash water for special analysis and disposal. 8.5 Management of Individuals After Contamination Assessment and Decontamination of Skin and Wounds • No injuries and no evidence of internal contamination: Individuals who have been decontaminated to the levels defined in Section 8.1 can be sent home with follow-up instructions prescribed by attending medical personnel. • Injuries and no evidence of internal contamination: Individuals who have sustained injuries but with no indication of radionuclide intakes through the injured tissues can be admitted to hospital without special restrictions. • Evidence of internal contamination with or without injuries: Individuals who show evidence of having inhaled or ingested radionuclides or have had intakes through the skin

8.6 DOCUMENTATION

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by absorption through wounds or by contaminated shrapnel and debris should be admitted to a hospital emergency department or a special radiation treatment facility (if available) for medical evaluation and treatment and for internal contamination assessment. The referral hospital should be notified as soon as it has been determined that the patient may require hospitalization. Initial decontamination procedures and preliminary evaluations at the incident site may have taken several hours but advance notice during this time will enable the hospital to mobilize its resources. The onsite personnel must transmit all their information to the hospital staff and remain in communication until all early diagnostic and treatment decisions have been made. Names and points of contact of work associates and supervisors must be obtained for reference, even after the exposed person(s) has been transferred to the hospital, in case further information on details concerning the incident is needed. 8.6 Documentation Record the decontamination procedures and results (see Appendix A, Table A.3.1).

Part C: Patient Management at Hospital 9. Stage 4: Patient Evaluation and Emergency Care (hospital)

Objectives • evaluate and treat patients with injuries and psychological distress; • evaluate patients for evidence of radiation sickness and begin treatment; • evaluate patients with radiation skin burns, contaminated wounds and radionuclide intakes for emergency treatment; and • confirm patients with possible internal depositions. See Figure 3.1 for flow of persons through all nine stages in the management of radionuclide contamination. 123

124 / 9. STAGE 4: PATIENT EVALUATION AND EMERGENCY CARE 9.1 General Issues in Initial Patient Evaluation Although most radiological contamination incidents do not involve potentially-hazardous levels of radiation exposure or radionuclide uptake, there have been a number of instances of radiation injury. Very-high levels of localized radiation of the skin from external sources or from deposited radionuclides can cause skin burns while exposure of the whole body can cause generalized skin burns and radiation sickness. Internal radionuclide contamination can also cause serious health problems. The route of uptake, whether inhaled, ingested or through the skin, plays an important role in the severity of the expected outcome and in treatment decisions. It is important that an early effort is made to determine the radiation dose received by the patient from both external and internal sources. 9.1.1

Medical Evaluation of Persons with Internal Contamination

Persons exposed to radiation and possibly contaminated with radionuclides require prompt treatment of medical and surgical conditions and an initial evaluation of radiation exposure. Since radiation-related illness requires hours to days to become clinically evident, hospital emergency personnel should triage individuals of a radiation-related incident using traditional medical and trauma criteria. Patients should be medically stabilized and then assessed for radiation injury based on clinical symptoms, estimated dose, radionuclide involved, and whether internal depositions have occurred (Section 10). Resource limitations may necessitate some differences in the level of care rendered in mass casualty situations. The “golden hour” (the first hour after a traumatic incident) is widely recognized by trauma surgeons to be that period of time during which the lives of severely-injured people may be saved, provided they are rapidly triaged to definitive treatment by firstresponse personnel. Patients in shock or near shock can die if not treated within the golden hour. In an incident involving the inadvertent or deliberate release of radioactive material, the presence of radiation must not interfere with rapid triage and removal of trauma patients from the field of injury. As described in Sections 7 and 8, internally-contaminated individuals should be evaluated to determine whether decontamination is advised. Removing contaminated clothing will eliminate 80 to 90 % of external contamination; soap and water should be the first approach to removing any remaining radioactive material. Irrigation and decontamination of wounds may be optimized using a tepid

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saline or water jet under mild pressure. Ambient radiation levels from contaminated wounds rarely exceed a few tens of μSv h–1 (a few mR h–1) and hospital personnel should be reassured that their dose from such an exposure will be insignificant or minimal. In the analysis of external irradiation incidents, simple radiation dose estimates are generally sufficient for medical decision making. In the particular case of lost sources, a point-source approximation (Section 3.2.3) is quite adequate for an initial estimate of exposure. When the patient enters the medical care system, adequate data are rarely available for a complete health-physics analysis of the incident. For a treating physician to provide proper treatment to a patient of a radiation incident, it is helpful to be able to diagnose radiation illness in a timely manner. In the clinical practice of medicine, the correct diagnosis of disease may be made in ~85 % of cases using only a well-collected and reviewed medical history (Koenig et al., 2005). However, since radiation effects are rarely taught in medical schools, physicians often do not include radiation injury in the differential diagnosis of common radiation-induced prodromal symptoms (nausea, vomiting and diarrhea). If early radiation injury or acute internal deposition is not a consideration, then rarely will the diagnosis be made correctly, at least in a timely manner. 9.1.2

Psychological and Behavioral Consequence Management After Radiation Incidents

Exposure to radiation from an external source or an internallydeposited radionuclide, whether from an unintended release or from a terrorist attack, can create uncertainty and fear. The psychosocial aspects of a large radiation incident or terrorist attack are examined in detail in NCRP Report No. 138 (NCRP, 2001a). Other useful sources of information about the psychosocial and behavioral aspects of radiation incidents include Becker (2001), DHHS (2009), and DHS (2003). In the aftermath of such an incident, members of the public will seek advice from both health-care providers and members of the scientific community whose job it will be to determine the extent of internal and external contamination and radiation exposure. Individuals are likely to experience feelings of vulnerability, anxiety, and lack of control. Internal depositions of radionuclides may be particularly anxiety-provoking because the patient essentially has little control over radionuclide decorporation and must rely on evaluation from the medical community. In addition, where there is a lack of consensus among experts, this can increase public fear and anger.

126 / 9. STAGE 4: PATIENT EVALUATION AND EMERGENCY CARE Concerns about public radiation exposure and/or radionuclide contamination may also cause some exposed persons to seek medical treatment, even when none is indicated. For example, there have been many cases at industrial sites where workers had unintended inhalation of actinides with minimal activity. Despite their familiarity with radioactivity, some workers have nevertheless requested treatment with DTPA and other medications for years after the incident, even when shown that further chelation therapy would provide little medical benefit. Individuals who have internal radionuclide depositions will generally fall into one of two groups: those who manifest behavioral changes immediately or shortly after contamination and those who may experience mental health impacts at a future time. If there is a release of radioactive material, public distress will be common and may manifest as anxiety, sadness, fatalism, anger, fear, difficulty sleeping, impaired ability to concentrate, disbelief, or other symptoms. Psychological distress after a radiologic incident may also manifest as nonspecific somatic complaints [a presentation sometimes referred to as multiple idiopathic physical symptoms (MIPS)]. Fortunately, for the vast majority of people, distress and psychological and behavioral symptoms related to the traumatic incident will diminish over time. For a smaller number of others, however, symptoms will persist, affect function at home and work, and may even result in psychiatric illness. In the early aftermath of a radiation incident, many individuals may fear that they have been contaminated and will misattribute signs and symptoms of autonomic arousal to radiation. In the longer term, individuals may present to primary care providers with multiple somatic complaints for which no etiology can be determined. A well-organized and effective medical response can instill hope and confidence, reduce fear and anxiety, and support the continuity of basic community functions. Mental health professionals should be an integral part of the teams planning and performing initial screening and triage. Where feasible, the establishment of an “Emergency Services Extended Care Center” (a term first developed by the Rush Chicago Medical Center) may offer an important means of assisting persons who remain fearful even after receiving negative findings. Reinforcing self-reliant behavior, fostering self efficacy, and providing information that can be used to protect oneself and one’s family will help alleviate distress. Finally, the psychological value of receiving appropriate medical countermeasures and information about methods of self protection can be substantial (Koenig et al., 2005). The long-term aspects of these issues are discussed in greater detail in Section 13.4.

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9.2 General Instructions for Emergency Department Medical Staff Guidance for emergency room medical staff does not differ greatly from guidance to medical staff at the incident site, except for the greater availability of medical resources that allow for more intensive procedures. Individuals exposed to releases of radionuclides are unlikely to be admitted to a hospital emergency room unless they have incurred serious injuries or serious internal depositions of radionuclides. This is true whether the incident occurs in a facility with radiation-safety and medical support or in mass contamination incidents involving large numbers of people. When persons likely to be contaminated with radionuclides are evaluated in hospital emergency departments, it is important, if at all possible, to have radiation-safety professionals available as well as the usual medical staff. The presumption may be that only those with serious injuries and/or obvious significant intakes of radionuclides will appear at hospital emergency departments, but in reality following a mass exposure incident, there will be many “worried well.” The report entitled Generic Procedures for Medical Response During a Nuclear or Radiological Emergency (IAEA, 2005b) is recommended as a source of further information. The following general guidelines are appropriate for most contamination incidents: • Contaminated or irradiated patients do not pose a health risk to medical responders except in extremely rare situations where the patients are heavily contaminated and the activity may become airborne or have highly-radioactive material imbedded in the skin. • Medical personnel should consider both medical and radiation issues, but critical medical concerns should be addressed first (Koenig et al., 2006). • Medical personnel should not be expected to advise pregnant women on the medical effects of their suspected radiation exposure until they have knowledge of the fetal radiation dose estimate. • Nausea and vomiting are signs of radiation sickness. However, unless exposed to a very-high-activity radioactive source, patients are unlikely to be exposed to sufficiently high doses of radiation to cause radiation sickness (Table 6.2 describes signs and symptoms of whole-body radiation exposure). • Except in cases of extremely-high exposure, radiation injury typically has a latency period of days to weeks. Skin burns,

128 / 9. STAGE 4: PATIENT EVALUATION AND EMERGENCY CARE vomiting, weakness, and other apparent radiation effects that appear after a short time may be caused by other factors (IAEA, 2005b). 9.3 Emergency Medical Management 9.3.1

Caring for Contaminated Individuals with Life-Threatening Injuries

A badly-injured patient is one at risk of losing their life, a limb, or their vision if treatment is delayed. In such cases, it is essential to attend first to life-threatening injuries. • Stabilize serious injuries and trauma first. If contamination is found around nose, mouth, or open wounds, or if high radioactive airborne concentrations are known or suspected to have occurred, consider the potential for internal contamination and sample before decontamination if possible (Section 10.3.1.1 has details on taking and interpreting nose swabs). • Decontaminate when patient is stable; follow contamination control measures (Section 5.3). • Consult with a radiation-safety professional (health physicist), radiation oncologist, or other radiation medicine expert for further evaluation and dose reconstruction. • Consider psychological (Section 9.1.2) and social needs of exposed individual, family and friends (Appendix E). This applies to all of the following categories of care in Sections 9.3 and 9.4. 9.3.2

Caring for Lightly Injured and Uninjured Contaminated Exposed Persons

Exposed Person is Contaminated and Lightly Injured • High radiation dose is most likely only to those nearest an incident. • If contamination is found around nose, mouth, or open wounds, or if high radioactive airborne concentrations are known or suspected to have occurred, consider the potential for internal contamination and collect samples (Section 10.3). • Decontaminate patients if possible, or implement contamination controls. • Treat minor injuries and other illness as appropriate.

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• Consult with a radiation-safety professional (health physicist), radiation oncologist, or other radiation medicine expert for further evaluation and dose reconstruction. Exposed Person is Contaminated but Uninjured • If contamination is found around nose or mouth, or if high radioactive airborne concentrations are known or suspected to have occurred, consider the potential for internal deposition and collect samples (Section 10.3). • Decontaminate if possible, starting with the face, and then the most contaminated other areas (Section 8.1.2). • Consult with a radiation-safety professional (health physicist), radiation oncologist, or other radiation medicine expert for further evaluation and dose reconstruction. 9.3.3

Caring for Persons Suffering from Radiation Injury

Only under rare circumstances will individuals contaminated with radionuclides incur significant whole-body irradiation and suffer serious early radiation effects. Examples are large intakes of soluble forms of 137Cs or 210Po. More information on these radionuclides appears in Section 12. However, it is important for emergency department physicians to be able to recognize such cases, because persons suffering early radiation effects will most likely appear at hospital emergency departments. Table 6.2 describes the signs and symptoms that result from acute, whole-body irradiation from an external source. The doses necessary to produce similar effects from internally-deposited radionuclides are substantially higher because they are received at much lower radiation dose rates. The doses persons receive in an incident occurring at a site where radiation monitoring is in place usually will be quickly estimated and the exposed individuals placed in one of the following categories. However, estimates of doses received by persons in a mass exposure incident at a site lacking radiation-monitoring data will usually not be known for several days. Therefore, assigning exposed individuals to one of the following categories will have to be tentative, based largely on signs and symptoms (Table 6.2). The following apply to all those suspected of receiving wholebody radiation: • stabilize serious injuries first; take appropriate contamination control actions;

130 / 9. STAGE 4: PATIENT EVALUATION AND EMERGENCY CARE • if contamination is found around nose, mouth, or open wounds, or if high radioactive airborne concentrations are known or suspected to have occurred, consider the potential for an internal radionuclide deposition and collect samples (Section 10.3); • decontaminate patient when stable; • perform assessment of external dose by obtaining a series of CBCs; • complete radiation dose assessment; consider observation with serial CBCs every 8 h; • treat minor injuries; • consult with a radiation-safety professional (health physicist), radiation oncologist, and radiation medicine expert for further evaluation and dose reconstruction; and • consider psychological and social needs of patient, family and friends. Exposed Person has Received Potentially-Lethal Radiation Dose: >6 Gy (600 rad) • This level of exposure is unlikely following the explosion of a radiological dispersal device (RDD) but may occur in case of a nuclear attack or when exposed to a high-activity radioactive source. • Nausea and vomiting should be taken seriously as indicators of radiation exposure. However, these symptoms can be caused by factors other than radiation (medical, psychological, trauma). If vomiting only occurs once, it is probably not radiation-induced. • For higher exposures, consider placing patient in reverse isolation • Patients exposed to >6 Gy (600 rad) will need intensive medical care due to profound immune suppression. Details of such care are provided in the article by Waselenko et al. (2004) and other similar publications. • Patients exposed to ~8 Gy (800 rad) will likely have received a lethal radiation dose. Exposed Person has Received a Serious Radiation Dose: 1 to 6 Gy (100 to 600 rad) • If the radiation dose is >1 Gy (100 rad), complete all surgical procedures within 2 d of exposure. • For higher exposures, consider placing patient in reverse isolation and managing as described above.

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Exposed Person has Received Mild Radiation Dose: <1 Gy (100 rad) • Doses in this range are unlikely to result in outward signs or symptoms. • Decontaminate patients if possible, or implement contamination controls. • Treat minor injuries and other illness as appropriate. 9.3.4

Caring for Persons Suffering from Radiation Burns

High levels of radiation exposure can damage the basal cells of the skin. This can lead to blistering (wet desquamation), peeling (dry desquamation), reddening (erythema), and loss of hair (epilation). Some symptoms (erythema, possibly itching) may appear transiently shortly after exposure, reappearing days or weeks later, accompanied by more severe symptoms depending on the level of exposure. Injury in small areas of the skin will normally heal by normal regeneration; larger-scale injury may lead to permanent discoloration, hair loss, fibrosis, tissue atrophy, ulceration, and/or necrosis of the affected tissue. Some of these effects and the radiation dose needed to produce them are summarized in Table 9.1. It is important to note that patients can have a high skin dose with little corresponding dose to the whole body or deeper lining organs and tissues [see NCRP Report No. 156 (NCRP, 2006a) for a detailed description of radiation burns and the healing process]. Erythema, ulceration, and other effects can be caused by chemical and thermal injury as well as by radiation, and it may be necessary to differentiate between these causes. TABLE 9.1—Radiation skin injury and associated skin dose (Gusev et al., 2001; IAEA, 1998b). Stage/Symptoms

Dose Range (Gy)

Time of Onset (d)

Epilation

>3

14 – 18

Erythema

3 – 10

14 – 21

Dry desquamation

8 – 12

25 – 30

Moist desquamation

15 – 20

20 – 28

Blister formation

15 – 25

15 – 25

Ulceration (within skin)

>20

15 – 25

Necrosis (deeper penetration)

>25

>21

132 / 9. STAGE 4: PATIENT EVALUATION AND EMERGENCY CARE • Many locations where radioactive materials are used do not contain sufficient concentrations to cause radiation injury if the sources are breached. Exceptions are high-activity radioactive sources such as those used in industrial radiography, radiation sterilization, well-logging, and those encountered in remediation of DOE former weapons production sites (Section 17). • In a laboratory setting, the risk of chemical and thermal burns is greater than radiation burns because most laboratories do not have radiation sources with sufficient activity to cause radiation skin injury (burns). • If the patient denies chemical or thermal contact, radiation should be suspected. • In a terrorist RDD attack, thermal injury from the explosion and subsequent fire is more likely than radiation skin injuries. Chemical and thermal exposures cause symptoms different from those caused by radiation and these can help identify the type of exposure. 9.4 Initial Treatment Decisions Individuals exposed to small levels of internal radionuclide depositions may not need any treatment. Since no treatment is completely free of risk, a benefit-to-risk decision must be made before embarking on an aggressive course of treatment. On the other hand, prompt action is most effective for individuals with significant internal depositions. However, undertaking mitigation procedures in the emergency department would be a rare occurrence. Generally such procedures would occur after the patient is admitted to hospital or on an outpatient basis. The following are possible exceptions. It should be noted that the use of interventional techniques to enhance the body’s natural elimination rate of the compound, or possibly block the uptake of the radionuclide in a particular tissue, may partially or completely invalidate the use of standardized model approaches to estimate the intake and dose. 9.4.1

Radionuclides in a Wound

Radionuclide contamination of a wound is a clear indication for a decontamination attempt. If not removed, the radionuclide may eventually be absorbed into the body, where it is metabolized and deposited in a target organ or tissue. NCRP Report No. 156 (NCRP, 2006a) provides detailed guidance. Case studies in Section 20 describe decontaminations of wounds in individuals and treatments and their outcomes for internal depositions of a number of radionuclides.

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In treating wounds contaminated with radionuclides, the first priority is to achieve homeostasis and take whatever medical and surgical measures may be indicated to preserve physiologic and anatomic function. Open contaminated wounds may allow rapid incorporation of radionuclides into the body, so they should be copiously irrigated with physiologic saline solution or sterile distilled water for several minutes. Depending upon the radionuclide, irrigation of the wound with a chelating agent such as DTPA may be considered. An effective irrigation solution is 1 g calcium diethylentriaminepentaacetate (Ca-DTPA) and 10 mL 2 % lidocaine in 100 mL of 5 % glucose solution or isotonic saline (NCRP, 1980; 2006a). If contamination with plutonium or other actinides is suspected, consideration also should be given to systemic chelation (IV, IM, or oral). Care should be exercised to avoid possible overdoses of chelating agents, since the amount absorbed when irrigating wounds is impossible to measure. Scrubbing action employed in cleansing the wound is more important than the cleansing agent used (NCRP, 2006a) (see Section 12 for further information on chelating agents). If contamination persists within the wound, surgical debridement may be necessary. When surgical exploration and excision of contaminated tissue/foreign material is necessary, it should be performed with assistance of a radiation-safety professional (health physicist) using a wound probe. The excised material should be saved for radioanalysis. There is no contraindication to the use of standard local or systemic anesthetic agents in managing these types of wounds. A wound containing radioactive material can be treated by primary closure once adequate decontamination has taken place as determined by monitoring the wound site and/or measuring the activity in the excised tissue and collected washes. Decontamination of intact skin usually requires only gentle scrubbing with soap and warm (not hot) water (Section 8). Aggressive scrubbing should be avoided because it can increase absorption of radionuclides through the skin. Hair can usually be decontaminated with soap and water. However, if this is not successful, it may be necessary to remove the hair. Hair should be clipped rather than shaved, to avoid disrupting the skin barrier and possibly enhancing absorption of the contaminating radionuclide. 9.4.2

Radionuclide Inhalation

Suspected inhalation intakes of a radionuclide should be evaluated by collecting nasal swabs (Section 10.3.1.1). If the nasal swab is positive for radioactive material, the individual’s nasal area may

134 / 9. STAGE 4: PATIENT EVALUATION AND EMERGENCY CARE be cleared by wiping or irrigating the area. Additional studies as described in Section 10.3.1 will be necessary to determine the body burden of internal depositions and necessity for treatment. However, there are some situations where prompt administration of medical countermeasures is recommended based only on positive indications of an intake. For example, if intakes of plutonium or americium are suspected (e.g., from surveys of the facial areas or by positive nasal swabs), DTPA should be administered immediately when it is most effective in hastening the removal of these elements from the body, before substantial deposition in tissues such as bone and liver has occurred. The safety and efficacy of DTPA has been well demonstrated (Breitenstein, 2003; IAEA, 1978; NCRP, 1980; Norwood, 1975) (Section 12.3.3). Specific drug therapy and lung lavage procedures are described in Section 12. Another example is when there is evidence of intakes of radioiodine either by inhalation or by other routes. In such cases prompt administration of KI will block uptake of radioiodine in thyroid tissue (Section 12.4.3). 9.4.3

Radionuclide Ingestion

Ingestion of a radionuclide may be considered as ingestion of a “poison.” The radionuclide will have a transit time through the GI tract prior to absorption into the blood stream (Sections 16 and 20). Prompt action may reduce the amount of radionuclide absorbed. Consult Section 12 for further guidance. GI absorption of radionuclides can be reduced by limiting the amount of time they remain in contact with the lining of the digestive tract. This can be achieved either by chemically or physically purging the intestines or by use of medications selected for specific elements (Section 12). Some medications will combine with radionuclides rendering them less available for absorption and more rapid elimination in the stool. Examples of such medications include the alginates and aluminum-containing compounds that tightly bind radioactive strontium. The following procedures can be used for reducing the GI absorption of radioactive substances. Hospital personnel should be advised by radiation-safety (health physicist) staff on the use of PPE in conducting these procedures and on steps to be taken to minimize contamination of the facilities and other personnel. 9.4.3.1 Gastric Lavage. This procedure may be useful in the first few hours after ingestion of a radionuclide for evacuation of material from the stomach. It would only be used in the highly unusual circumstances where the intake of a large quantity of radionuclide

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is suspected that might pose a significant threat to health. Gastric lavage should be performed in a standard manner until the stomach washings are relatively free of radioactive material. The washings should be saved for activity measurements in a laboratory. If the procedure is unsuccessful for any reason, consider the use of purgative drugs. 9.4.3.2 Emetics. The use of emetics is controversial. The American Academy of Pediatrics recommends that ipecac (an emetic) not be used routinely at home, and that caregivers consult with their local poison-control center. If an emetic is used, radioactive material obtained should be sent to a laboratory for analysis. 9.4.3.3 Purgatives. Purgatives or laxatives are classified generally as irritants, bulk-forming substances, lubricants, and wetting agents. Castor oil, cascara, and senna are examples of irritants. Slower acting drugs, including bulk-forming laxatives and wetting agents, are not generally appropriate. Use of enemas or colonic irrigations may be valuable in reducing the residence time of radioactive materials in the colon. Selection of a purgative should include the consideration of speed of action. Furthermore, it may have special properties by which the purgative itself may produce a less soluble compound of the involved radionuclide. Magnesium sulfate is an example of a saline cathartic that may produce relatively-insoluble sulfates with some radionuclides and thus reduce absorption. The fact that magnesium sulfate has been used should be indicated to the radiochemist who will analyze the fecal sample since it introduces complications in the analytical procedure for certain radionuclides. Saline cathartics usually act within 3 to 6 h. Their action depends on the poor absorption of cations, anions or both. A hypertonic solution is produced in the intestine and water is attracted from the intestinal mucosa. All laxatives are contraindicated in the presence of abdominal pain of undetermined etiology, abdominal obstruction, or an acute surgical abdomen. Adverse reactions with laxative use may include dehydration, cardiac irregularities, enteritis, dyspnea, syncope, rash, and loss of electrolytes, especially of potassium, which can cause weakness. These are generally associated with laxative abuse. No studies are noted in the literature on the usefulness of activated charcoal to decrease radionuclide absorption from the intestine, but it should be a potentially-useful procedure. Until studies have delineated its usefulness, activated charcoal ingestion may be considered an untested alternative to ion exchange resins (NCRP, 1980).

136 / 9. STAGE 4: PATIENT EVALUATION AND EMERGENCY CARE 9.4.4

Clinical Decision Guides

A new operational quantity, Clinical Decision Guide (CDG), is introduced in this Report to aide practitioners in making decisions about treatment of persons that have internally deposited radionuclides. Full details of the CDG concept and its use related to radionuclide-exposed persons are given in Section 11. 9.4.5

Specific Drug Decorporation Therapy

Medical mitigation of the effects of internal depositions will rarely be necessary in the emergency department. As noted above, an exception is when nasal swabs or surveys of nose and mouth give strong evidence of an intake of an actinide such as plutonium. In such cases the chelating agent, DTPA, should be given as promptly as possible. However, although starting treatment within a few hours is advisable, chelation with DTPA has averted significant dose in some patients even when treatment was started weeks later.6 Decorporation with chelating and other agents generally consist of a series of administrations over days, weeks, and sometimes months. Section 12 provides guidance on decorporation therapy. 9.4.6

Algorithm for Medical Management of Internal Depositions

Radiation Event Medical Management (REMM) is a comprehensive web portal provided by DHHS for the diagnosis and treatment of radiation injury. The website has a useful algorithm for the management of internal depositions (DHHS, 2009). 9.5 Medical Information Checklist The questions that follow can be used by the attending physician at the hospital for obtaining historical information to assist in the early management of persons contaminated with radioactive material. The best information in industrial cases can probably be obtained from plant personnel, such as a health physicist or an occupational physician familiar with the plant and incident details. Information collected in the emergency department will supplement the information obtained onsite (Section 6). Appendix A contains sample forms that may be used to document radionuclide contamination incidents. 6Wiley,

A., Jr. (2008). Personal communication (Radiation Emergency Assistance Center/Training Site, Oak Ridge, Tennessee).

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• When and where did the radionuclide release occur? • What are the circumstances of the incident and what are the most likely pathways for contamination? • How much radioactive material is involved potentially? • What injuries have occurred? • What underlying health problems may be present besides the radionuclide contamination? • Are toxic or corrosive chemicals involved in addition to the radionuclides? • Have any treatments been given? • What radionuclides now contaminate the patient? • Where is the contamination located? • What are the radiation measurements at the surface? • What information is available about the chemistry of the compounds containing the radionuclides? • What activity measurements have been made at the site of the incident (e.g., air monitors, smears, fixed radiation monitors, nasal-swab counts, and skin contamination levels)? • What decontamination efforts have already been attempted and how effective have they been? • Have any therapeutic measures, such as administration of blocking agents, chelating agents, or isotopic dilution, been given? • Was the individual also exposed to penetrating radiation from an external source and, if so, is dose information available? • If onsite decontamination occurred, was clothing saved? Contamination still on clothing can be subjected to radiation energy spectrum and chemical form analysis and examined for particle size. • What excreta have been collected? Who has the samples? What analyses are planned? When will results be available?

10. Stage 5: Internal Contamination Assessment (hospital)

Objectives • determine routes of entry of radionuclides into the body; • identify the radionuclides and their physical and chemical forms; and • determine radiation doses using in vivo and in vitro bioassay procedures. See Figure 3.1 for flow of persons through all nine stages in the management of radionuclide contamination. 10.1 Preliminary Assessment Activities Identification of exposed persons who may have had intakes of radionuclides and are, thus, internally contaminated will usually have occurred onsite by radiation-safety personnel (Section 6). After decontamination, persons with internal contamination should be admitted to an emergency facility for a complete assessment of the intake. Triage documentation (Section 6) describing the contamination incident, radionuclides, possible routes of intake, and all other pertinent information should accompany the contaminated person for use by medical staff and health-physics trained professionals. Prompt assessment of possible internal contamination is essential to initiating appropriate treatment. Internal 138

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depositions can occur via inhalation, ingestion, absorption through skin or wounds, or penetration of the skin by contaminated shards, shrapnel or debris. Table 10.1 describes procedures that should occur and the samples that should be taken as early as possible after the patient is admitted to the emergency room. The results can guide more rigorous assessments using advanced methods, as needed.

TABLE 10.1—Preliminary dose-assessment procedures. Exposure Pathway

Assessmenta

Skin or clothing contamination

Measure radiation dose rate from contamination

Inhalation (nasal and/or oral)

Nasal swabs Survey facial area with portable instrument (alpha and beta/gamma) Obtain count rate from lungs using portable instrument (gamma only)

Ingestion

Facial survey with portable instrument

Skin absorption

Whole-body survey with portable instrument. Swabs of skin and all body orifices, ears, etc.

Absorption via a wound

Obtain count rate from wound using portable instrument

Injection (e.g., contaminated shrapnel)

Obtain count rate at site of injection (or wound) using portable instrument

Whole-body radiation exposure (can occur from intakes of radionuclides that deposit throughout the body as well as from external sources on the skin)

Calculate dose using lymphocyte depletion or cytogenetic assays after 24 h

a In all cases of possible internal depositions, collect urine, feces, and possibly blood. The latter is rarely useful in assessing radionuclide intakes but can be used to estimate radiation dose using cytogenetic biodosimetry.

140 / 10. STAGE 5: INTERNAL CONTAMINATION ASSESSMENT 10.2 Information About the Contaminating Incident For physicians to make decisions regarding appropriate medical treatment, estimates of the dose of radiation absorbed by both individual organs/tissues and the whole body are required for each person suspected of internal contamination or exposure to a high external-radiation field. These dose assessments will require a minimal set of information on the specifics of each patient admitted to the hospital. For intakes, establishing the location and time spent by the individual at the contamination scene, suspected route or routes of contamination, and when possible, the likely physical and chemical form of the radionuclides involved in that contamination, are key pieces of needed information. For high exposures to external radiation, dose rate, time spent near the source, and location of the individual relative to the source, are also key pieces of needed information. For collimated beams, it is helpful to know where the beam intersected the body and its cross-sectional area. Once these features of the exposure are established, decisions can be made on the type(s) of biological samples to be collected and their collection frequencies. Most of this information should be obtained by the onsite radiation-safety professional (health physicist). When that does not happen, patients should be interviewed at the appropriate time (when stable) with the information relayed to the radiation-safety personnel responding to the incident and those assigned to perform the dose assessment. Appendix A contains sample incident documentation forms. For more detailed information the reader is referred to the Scientific and Technical Bases: • Section 19, Instrumentation to Measure Radioactive Contamination; • Section 20, Dosimetric and Case Studies for Selected Radionuclides; and • Section 21, Dose-Assessment Methodologies.

10.2.1 Location of the Individual and Time of Exposure As outlined in Section 5, surveys of both airborne activity and surface contamination should be made at the incident site. These activity data may then be used to reconstruct both external exposures and possible external and internal radionuclide contaminations. By knowing a person’s locations following the original radionuclide release (e.g., explosion, spill, leakage, etc.) and the

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length of time spent in each place, health physicists can determine who may be at the greatest risk for radiation-induced adverse health effects. Known sites of environmental contamination, their activity levels, and radionuclides present can be plotted on a map and when appropriate medically, individuals should be interviewed to document their location at the time of the contamination incident. 10.2.2 Establishing the Route of Exposure Inadvertent internal contamination with radionuclides may occur via three primary routes: • inhalation of aerosol particles containing radionuclides (either through the mouth or nose); • ingestion of materials from hand and face surface contamination, as well as consumption of contaminated food or water; or • direct absorption into blood following radionuclide contamination of wounds, cuts, or abrasions or from penetration of the skin by contaminated shards, shrapnel or debris. Even when internal activity is known to be present (as seen through the analysis of urine samples or direct counting with photon detectors) it is important to determine how that activity originally entered the body. If the incident involved the release of radioactive material into air within the vicinity of the individual (e.g., a ruptured steam pipe in a nuclear facility or detonation of an RDD in a crowded metropolitan area), then inhalation is a likely route of entry. Ingestion routes may be more likely following external contamination of skin and clothing with the exposed individual wiping their face, brow or mouth prior to onsite decontamination procedures. If skin lesions are present on the individual, one must additionally consider direct blood uptake via the wound site (NCRP, 2006a). Personal interviews with the exposed individual, in combination with physical examination, whole-body counting, and/or bioassay analyses, are thus important to reconstruct routes of exposure and their relative contribution if more than one route is suspected. Some questions to ask are: • Are there any wounds? Are they contaminated? • Is there nasal contamination? If so, consider probable lung deposition. • Is there oral contamination? If so, consider inhalation and ingestion intakes.

142 / 10. STAGE 5: INTERNAL CONTAMINATION ASSESSMENT When available, air samples measuring positive for radionuclides are strong indicators of possible inhalation intakes by persons in the area. However, only in very extreme situations can the inhaled amounts approach levels that would cause deterministic effects. Relatively-high air concentrations of radionuclides would have to be sustained for many minutes or even hours for such intakes to occur. For example, the second column in Table 10.2 gives estimates of the air concentrations of several radionuclides that would have to be breathed for 10 min to result in intakes that would approximate threshold radiation doses for deterministic effects. The third column gives estimates of the air concentrations of the same radionuclides that would have to breathed for 10 min to result in an effective dose of 0.25 Sv (25 rem) (Section 16.7). An effective dose of 0.25 Sv (25 rem) is the value selected for 1 CDG, for stochastic effects (Section 11). It can be seen that these tabulated air concentrations are very high and it is unlikely that many individuals will be exposed to these high levels. 10.2.3 Radionuclide Identification and Physical and Chemical Form In most cases, energy spectral information acquired during whole-body counting, partial-body counting, and/or bioassay measurement will allow for energy identification of the radionuclide(s) involved in the contamination incident. Additional information, particularly for inhalation routes of intake, should be gathered regarding the physical and chemical form of the radionuclide. Aerodynamic properties, such as the physical size and shape of aerosol particles containing the radionuclides, play a significant role in determining where inhaled particles will be deposited initially within the respiratory tract (e.g., different parts of the upper respiratory tract or within deeper airways of the lungs). The chemical form may be even more important and is generally more likely to be learned. The chemical form determines where in the body the radionuclide will be deposited, the mechanism and rate for its transport, and the mechanism and rate of its excretion from the body. Preferably, this information should be gathered from the managerial staff of the facility in which the incident occurred. In the case of a radiological terrorist incident, related information should be collected by the onsite response team if possible. Patient interviews, however, would be appropriate as a source of supplemental information, especially following an occupational incident involving a worker knowledgeable about radiological materials.

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TABLE 10.2—Estimates of the concentrations in air [MBq m–3 (μCi m–3)] of several radionuclides that would have to be inhaled for 10 min to achieve intakes sufficient to produce deterministic effects or give effective doses of 0.25 Sv (25 rem) (Section 16.7). Radionuclidea

90 SrCl2 (Type F) 131I (Vapor) 137CsCl (Type F) 144

CeO2 (Type S) 210PoCl 2 210 PoCl4

Air Concentrations Required to Cause Deterministic Effectsb,c

Air Concentrations Required to Result in an Intake of 1 CDG Leading to an Effective Dose of 0.25 Sv (25 rem)c

2,600 (70,000) Bone-marrow depression

51

(1,400)

30 (800) Hypothyroidism

76

(2,100)

8,000 (220,000) Bone-marrow depression

350

(9,500)

3,700 (100,000) Pneumonitis

52

(1,400)

1,900 (51,000) Bone-marrow depression

0.67

(18)

PuO2 (Type M)

40 (1,100) Pneumonitis

0.049

(1.3)

239PuO

40 (1,100) Pneumonitis

0.18

(4.9)

40 (1,100) Pneumonitis

0.57

(1.5)

or

(Type M) 238

2

(Type S) 241

AmO2 (Type M) aThe

radionuclides shown here are used as examples to demonstrate the levels of airborne activity required to cause serious health concerns. Assumed a breathing rate of 1.2 m3 h–1 of unfiltered air for an adult and a lognormal particle-size distribution with AMAD = 5 μm. b Deterministic effects expressed within two to three months are given for the particular radionuclide. c Calculations based on a breathing rate of 1.2 m3 h–1 of unfiltered air by an adult, 5 μm AMAD particles, and a total deposition of 82 % (ICRP, 1994a).

10.3 Bioassay The bioassay methods used to assess internal contamination are categorized as either direct (in vivo) measurements or indirect (in vitro) measurements. In vivo measurements involve placing the patient near a radiation detector and measuring the radiation

144 / 10. STAGE 5: INTERNAL CONTAMINATION ASSESSMENT emitted from that person. In vitro measurements involve a radiochemistry laboratory analysis of material removed from or excreted by the patient, typically in urine or feces. Depending on the radionuclide and circumstances of intake, one type of measurement may be preferred over another or combinations may be appropriate. Facilities that routinely deal with radioactive materials (where employees are at risk for internal contamination) will typically have radiation-safety (health-physics) staff knowledgeable in radionuclide contamination measurements, and their interpretation. Likewise, occupational medicine facilities responsible for providing care to the employees of radionuclide handling or processing facilities may have special training, procedures, and experience in support of radionuclide intakes. Broader medical support facilities (e.g., community hospitals) may not have the necessary experience for obtaining and interpreting radiobioassays due to the extremely infrequent nature of their involvement in such exposure incidents. REAC/TS (Appendix G) can provide consultation with regard to bioassay methods and access to facilities having radiobioassay capability. The following discussion provides a summary of types of bioassay, their applications, and limitations. More detailed discussions of measurement techniques and interpretations can be found in the Scientific and Technical Bases. Section 19 describes bioassay instrumentation and analytical procedures. Section 20 is a compilation of biokinetic and dosimetric information on important radionuclides of 24 elements. Section 21 provides information on assessment of doses from bioassay data. Other useful sources include ANSI/HPS (1996), IAEA (2004a), and ICRP (1997). 10.3.1 Indirect (in vitro) Bioassay Monitoring In vitro bioassay monitoring includes analysis of nasal swabs, urine samples, fecal samples, blood, and tissue specimens. Approved protocols should be followed in collection of these samples. When collected in a hospital, these special protocols should be communicated to nurses and floor attendants. 10.3.1.1 Nasal Swabs. Nasal swabs can be a useful indicator that an inhalation intake may have occurred, however their interpretation is subject to very large uncertainties. While there is some debate about their value as a dose estimation tool, there is agreement that they can provide indication of a significant inhalation exposure. Two alternate methods for collecting and analyzing nasal swipes are described, and guidance in their interpretation is provided.

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Nasal swabs for radionuclides should be obtained from any patient suspected of inhaling radioactive material. They should be collected as soon as the patient’s condition permits, and prior to showering or washing, preferably within 30 min of the contamination incident and in a radiological clean area. This procedure must be performed early to obtain accurate results before an individual puts a finger in, or blows his/her nose. The exterior passages of the nose also can clear rapidly due to exhalation and nasal drainage. The typical nasal-swab procedure is to use a slightly moistened cotton tipped applicator and gently rotate the swab around the accessible surface of the nostril. A separate swab should be used for each nostril to prevent cross contamination of either the swabs or the nostril. A variation on this technique is to wrap a small (two inch) gauze pad or a 47 mm air filter paper (common in occupational exposure settings) around the applicator. The use of the gauze pad or filter paper allows a flat geometry, which is particularly suited for counting for alpha activity that can be shielded by a thin film of water. Each swab should be put into its own container (e.g., test tube, envelope, or bag), labeled with the subject’s name, collection time, date and location, and sent to an appropriate measurement laboratory. Swabs may be measured directly with a survey meter or a contamination smear counter. Swabs to be counted for alpha activity may require drying under a heat lamp to evaporate water film that can shield the alpha particles. Guilmette et al. (2007) have described a similar nasal-swab protocol specifically used for alpha (notably plutonium) contamination. Following swabbing, the cotton tip is trimmed from the applicator, placed in a liquid scintillation vial, and directly counted by liquid scintillation. Interpretation of the measurements of nasal swabs is subject to large uncertainties. The presence of activity on swabs, particularly if the results are similar for both nostrils, is presumptive evidence of inhalation of the radionuclide. High values in one nostril with much lower values in the other are suspect for contamination by means other than inhalation (e.g., rubbing one’s nose with a finger). The absence of measured activity in the nostrils cannot definitely rule out inhalation as a route of internal contamination. If the subject is a mouth-breather, either naturally, due to allergies or sinus congestion, or doing heavy work, nasal swabs may not be representative. The extrathoracic region of the respiratory tract is also subject to very rapid clearance, so if swabs are obtained hours after exposure, normal clearance may have already effectively eliminated radioactive material deposited in the nose.

146 / 10. STAGE 5: INTERNAL CONTAMINATION ASSESSMENT Opinions on the extrapolation of nasal-swab results to intake and dose range from the qualitative to quantitative. However, there is universal agreement on the importance of subsequent bioassay measurements for definitive intake and dose calculations. A ruleof-thumb used by some health physicists and physicians in evaluating nasal swabs after possible plutonium (alpha) contamination is that a value >80 Bq (~2 nCi) indicates a possible serious contamination, while results <1 Bq (~30 pCi) suggests only a possible low order of contamination. Guidance for occupational exposure at the DOE Hanford Site suggests that dose reduction therapy may be warranted for inhalations of transuranic radionuclides (e.g., 238Pu, 239 Pu, 241Am) based on nasal swabs >20 Bq (~0.5 nCi) in each nares, or 1,700 Bq (46 nCi) for beta activity, assuming 90Sr or 137Cs (Carbaugh, 2007). Mansfield (1997) offered a rough rule-of-thumb that the combined activity of both nasal swabs total ~5 % of deep-lung deposition, using the respiratory tract model of ICRP Publication 30 (ICRP, 1979) as a reference. Experience has shown this to be generally a conservatively high estimate of pulmonary deposition, useful for initial estimates pending bioassay results. Based upon the more recent Human Respiratory Tract Model of ICRP Publication 66 (ICRP, 1994a), together with results of a controlled study on humans, the combined activity of both nasal swabs is estimated to represent ~5 % of the total amount inhaled for 5 μm AMAD aerosols (Section 16.4.1.2). This value of 5 % is used in this Report to interpret nasal swabs for comparison with CDGs (Section 11 and Table 11.1). Guilmette et al. (2007) provided more quantitative guidance for 239 Pu based on experience at Los Alamos National Laboratory. In an extensive comparison of early nasal swabs analyzed by liquid scintillation with the final dose estimates, the value of 0.8 mSv Bq–1 (3 mrem pCi–1) was established for preliminary dose estimates based solely on nasal-swab data shortly following intake. Because of the highly variable and rapidly changing nature of nasal retention, NCRP is not recommending the use of nasal-swab results as a sole criterion when making decisions concerning dose intervention among members of the public following a radiation mass casualty incident. 10.3.1.2 Urine Bioassay. Urine bioassay is a versatile technique for measuring a wide range of systemically excreted radionuclides. Collection of urine is relatively straightforward, convenient, and easy to manage. Several example protocols for collection of urine to be used for dosimetry purposes are shown in Table 10.3. Following collection, the sample must be analyzed by a radioanalytical procedure appropriate for the radionuclide(s) of interest.

TABLE 10.3—Indirect (in vitro) sample protocols. Sample Protocol

Explanation

Application

Urine, single void, or “spot”

Collect the total volume of a single voiding of urine. Typical volume; 50 to 300 mL. The first urine voiding after an intake is likely to be nonrepresentative of the intake, because it will include urine accumulated in the bladder prior to the intake. Thus, it is recommended that the bladder be emptied as soon as practical after an intake, and the second voiding (best collected 4 h after the contamination incident) be used for intake assessment.

Provides an initial order-of-magnitude estimate of exposure based on an excretion model. Particularly useful as a “first indicator” of systemic uptake for plutonium or americium. It can also be used to estimate the initial efficacy of DTPA therapy for plutonium or americium. For many applications, single void samples will require some form of normalizing to give total daily excretion.a

Urine, partial day (see also urine ~24 h)

Collection of all urine voided in a 24 h period. The specific 24 h interval should be noted.

Preferred urine sampling protocol for most intake and internal dose assessments, because the least uncertainty is associated with its collection.

10.3 BIOASSAY

Urine, total 24 h

Usually used on the day of an incident when a person is given a urine sample kit to take home for rest-of-day sample collection. Extrapolation of results to simulated total daily excretion is usually required.

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Sample Protocol

Explanation

Application

Urine, ~24 h

Urine sample is collected under a standard protocol and normalized to an approximate 24 h total using a normalizing basis such as duration of collection interval, volume of sample relative to a 24 h reference volume, creatinine excretion, or specific gravity.

Often used for routine monitoring programs or special bioassay following a potential intake. Requires a normalizing protocol for interpretation using most biokinetic excretion models.a

Feces

Collect a single fecal voiding. Should know time interval represented by voiding.

Used to confirm an intake, identify isotopes, isotope ratio information, differentiate soluble from insoluble materials, and estimate intake and dose.

aNormalizing may be done by sample interval time, sample volume relative to reference excretion rate, creatinine, specific gravity, or a combination of such parameters, as discussed in the text.

148 / 10. STAGE 5: INTERNAL CONTAMINATION ASSESSMENT

TABLE 10.3—(continued)

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The subject should wash his/her hands prior to sample collection to minimize the potential for external contamination of the sample during collection and handling. Urine should be collected in clean containers that can be tightly sealed following sample collection. In the absence of standard sample collection containers, any clean plastic or glass liquid-tight container may be used. Samples must be adequately identified with personal identification (name, tracking number, etc.) and a sample date. The sample collection time should be noted if the sample is collected within 4 to 8 h of the contamination incident. Samples collected within 4 h of an internal contamination may not be representative of a true systemic uptake. Less than 4 h is an insufficient amount of time to allow metabolic processes to produce a representative sample: results may reflect dilution by urine stored in the bladder prior to the contamination incident. Consequently, it is recommended that samples be collected after at least 4 h. The sample size required for radiobioassay is variable, and can range from a few milliliters for tritium to total 24 h samples for actinides such as plutonium or americium. Ideally, total 24 h urine samples are preferred for intake and dose assessments, because most of the biokinetic models used for bioassay data interpretation are based on daily excretion rates. However, total 24 h urine samples are often neither practical nor convenient, and an approximate 24 h urine result may be obtained by normalizing a sample result based on one of the methods described in the Table 10.3 footnote. Normalization by sampling interval involves collecting all urine voided in a specified time interval [e.g., 30 min before retiring at night through 30 min after rising in the morning for two consecutive nights (NCRP, 1987)]. Medley et al. (1994) identified a potential bias of up to a factor of two for this method. Provided the sample is collected properly, a total or simulated 24 h urine sample result is used as is; no further normalization is done. A proper 12 h sample result would be normalized by doubling the result. Normalizing by sample volume usually assumes a reference excretion rate of 1,600 or 1,200 mL d–1 [adult male or female, respectively (ICRP, 2002b)], and adjusts the sample result based on the ratio of sample volume to reference volume. Normalization by creatinine or specific gravity has been suggested (Anderson et al., 1995; Duke, 1998; Karpas et al., 1998; NCRP, 1987; NIOSH, 1974). However, various studies suggest that normalization by these methods (creatinine or specific gravity) does not provide any improved confidence in the result over normalization by time or volume (Boeniger et al., 1993; Graul and Stanley, 1982; Jackson, 1966; Kim, 1995).

150 / 10. STAGE 5: INTERNAL CONTAMINATION ASSESSMENT 10.3.1.3 Fecal Samples. Fecal samples collected for radioanalysis to support dose assessments need to be total voidings, not simply stool smears or swabs. Such samples are useful for confirming radionuclide intake, identifying specific radionuclides, and differentiating between soluble and insoluble forms. They are a useful tool for investigating the ingestion of radionuclides, but are especially useful for assessing the inhalation intake of an insoluble radionuclide-containing material and its related internal dosimetry. Because of the time required to process fecal samples, they are not usually pertinent to initial screening assessments. Fecal samples should be collected in clean, air-tight, resealable plastic containers (e.g., commercial fecal sample kit or containers similar to one pint to two quart plastic ice-cream containers). Samples must be adequately identified with a personal identification (name, tracking number, etc.) and a sample date, and preferably the time or time interval since last voiding. Fecal voiding patterns are highly variable from person-toperson and day-to-day. Ideally a fecal sample should represent a total 24 h period; however such collection is often neither practical nor convenient. Fecal excretion for adults is noted by ICRP (2002b) to range from 50 to 500 g d–1, with a recommended reference value of 150 g d–1 for an adult male and 110 g d–1 for an adult female. Note that these values represent excretion “per day,” not excretion “per bowel movement.” When a single bowel movement is collected, it is generally interpreted as representing excretion for 1 d. Normalization to reference values is suggested if sample mass is low. If total accumulated fecal excretion over a time period was requested and there is no apparent reason to suspect that total excretion was not provided, then all sample results should be used as is, without regard for the mass of individual samples. If excretions were missed during the time period, then normalizing the sample mass to the reference mass is recommended by NCRP. 10.3.1.4 Blood Samples. Blood smears of a wound may be suitable for identifying initial wound contamination and for making crude screening decisions concerning therapy. For intermediate and highenergy beta/gamma emitters, lack of detection of radioactive material in a direct wound count and blood smear would suggest that dose intervention therapy is not indicated. By contrast, alpha emitters can be largely masked by the wet environment of a wound; lack of detection of alpha activity should not be construed as an absence of contamination. Taking a blood smear and drying it under a heat source (e.g., lamp) can allow detection of alpha activity using portable survey instruments (e.g., thin end-window GM detectors

10.3 BIOASSAY

/ 151

or alpha scintillator). Positive blood smears are an indication of wound contamination and may indicate potential systemic uptake. Blood smears (or skin breaks) containing >1.7 Bq (~0.05 nCi) alpha activity or 330 Bq (~9 nCi) beta/gamma activity suggest that dose intervention therapy might be warranted (Carbaugh, 2007). Collection of venous blood samples is indicated for obtaining lymphocytes, in particular, as baseline for clinical monitoring. Venous blood samples are also indicated for performing cytogenetic dosimetry evaluation following relatively-high dose [nominally 0.25 Sv (25 rem)] gamma- or neutron-radiation exposures. For cytogenetic analysis, 10 mL of peripheral blood should be collected in a lithium heparinized vacutainer tube. If dried heparin is used, it is important that the blood be properly mixed by inverting the tube several times. Samples should be placed in contact with a coolant pack in an insulated box. The samples should not freeze (IAEA, 2005b). Blood samples are not likely to be an effective medium for radionuclide analysis as part of an intake or internal dose assessment. The concentration of radionuclides of concern in circulation at any given time will likely be too small for detection by field instrument surveys and likely also to be too small for ready detection by typical radiochemical methods. Blood samples do not require any special considerations in collection beyond normal clinical standard precautions. They do not pose any radiological hazard to attending personnel. Blood samples may require the addition of an anticoagulant such as ethylenediaminetetraacetic acid (EDTA) at the time of collection to ensure uniformity in aliquots removed for duplicate analysis. 10.3.1.5 Tissue Specimens. Tissue excised from a contaminated wound, or material removed by wound debridement can be analyzed to determine the type and amount of radionuclide. Such information is important for assessing the total intake by a wound pathway. A specimen should be taken as soon as possible, with as little trauma to normal tissue as possible. A 3 mm skin patch is often useful in this regard. In an anatomically sensitive area such as the hand, a specialist should be consulted, such as a hand surgeon, if possible. Such specimens should be retained until a determination of their need and importance for the assessment process is made by appropriate medical or health-physics personnel. 10.3.2 Direct (in vivo) Monitoring Direct monitoring includes whole-body counting, chest (lung) counting, and special organ or tissue counting. Summary information about counting systems and their application is provided below and again in more detail in Section 19. Further information

152 / 10. STAGE 5: INTERNAL CONTAMINATION ASSESSMENT and discussions on these topics can be found in ICRU Report 69 (ICRU, 2003). 10.3.2.1 Whole-Body Counting. Whole-body counters measure high-energy photons (>200 keV) emitted from within the body. The most common detectors used in whole-body counters are either sodium iodide or germanium detectors. Low backgrounds are provided by the counting configuration, which may utilize a highdensity counting chamber or a shadow shield composed of iron or lead. Counting configurations include standing, sitting in a constant geometry chair, or reclining on a fixed or moving bed. Standing counts tend to be relatively short (e.g., 3 min) and correspondingly less sensitive than sitting or reclining counts which may run 10 to 20 min. Scanning whole-body counts (using a moving bed or moving detector) may be capable of providing information about radionuclide distribution in the body, as well as quantifying the amount present. Whole-body counters are typically found in large nuclear facilities (power reactors, nuclear material processing facilities, national laboratories, and specialized research facilities). Hospital gamma cameras may be capable of providing information similar to whole-body counters, although most have a more limited range of measured energies. 10.3.2.2 Chest (lung) Counting. Chest counting is typically used to measure low-energy photons (<200 keV), Most systems use germanium, sodium iodide, or phoswich [thin, dual NaI(Tl) and CsI(Tl) detector] detectors (Table 19.1) in a heavily shielded room. The subject is placed in a sitting, reclining or supine position. Counting times may range from 10 min to an hour or more. Actinides are of primary interest in chest counting, and chest counting facilities are usually found in national laboratories, actinide processing facilities, and specialized research facilities. Because the chest count can also include contributions from the skeleton (ribs, sternum and vertebrae), any measured results may require correction to give a more accurate estimate of the existing lung burden. Lacking correction, the measured chest count may erroneously be assumed to represent a lung burden. Chest count results can be highly dependent on chest-wall thickness, and chest-wall thickness is usually estimated for subjects by either a height-to-weight ratio or by measurement using ultrasound techniques. 10.3.2.3 Counting of Particular Organs or Tissues. The estimated radionuclide content of a particular organ or tissue can be of interest for routine monitoring (in the case of the thyroid) or as part of

10.4 INTAKE AND DOSE ASSESSMENT

/ 153

a special investigation following significant intakes (notably plutonium or americium). Thyroid counts are common within facilities handling radioiodines, and are typically obtained using sodium iodide or germanium systems. Counts of 241Am in the skeleton and liver are usually limited to special investigations following intakes of 239Pu or 241Am. Determining the radionuclide content of a specific organ or tissue can be quite complex; interference from radionuclides deposited in other tissues can falsely increase results. For example, measurement of 241Am in the lungs can be biased by 241 Am systemically translocated to the skeleton (especially the sternum and vertebrae) and the liver. Discussion of corrections for such interference is beyond the scope of this Report. 10.4 Intake and Dose Assessment The determination of intake and internal dose is an important outcome of the patient assessment and management. The process for calculating intakes and internal dose from bioassay measurements are addressed elsewhere in this Report. Section 19 describes instrumentation for in vivo and in vitro measurements. Section 20 provides dosimetry evaluations for radionuclides of 24 elements and Section 21 provides an overview of dose-assessment methodologies. The following table, Table 10.4, gives absorbed and effective dose7 coefficients for intakes of radionuclides (airborne radionuclides are assumed to have a lognormal particle-size distribution with a 5 μm AMAD unless otherwise specified). The data are taken from Section 20 which provides detailed metabolic and dosimetry modeling information, the basis for the dose coefficient calculations.

7The term effective dose, as used in this Report for internally-deposited radionuclides, always means committed effective dose calculated over a 50 y period beyond the radionuclide intake for adults and from intake to 70 y of age for intakes by children.

Absorbed Dose (30 d) Radionuclidea

Half-Lifeb

Main Radiationsc

Method of Measurementd

Exposuree Tissues

Effective Dose (50 y)

(Gy Bq–1)

(rad μCi–1)

(Sv Bq–1)

(rem μCi–1)

227Ac

21.8 y

αβγ

NS, IVC, U, F

Inh Type M

Lung

8.2 × 10–7

3.0 × 100

4.8 × 10–5

1.8 × 102

241

Am

432 y

αγ

NS, IVC, U, F

Inh Type M

Lung

6.3 × 10–7

2.3 × 100

2.7 × 10–5

1.0 × 102

252

Cf, D

2.65 y

αγn

NS, BC, U

Inh Type M

Lung

9.7 × 10–7

3.6 × 100

1.1 × 10–5

4.1 × 101

Red marrow

1.3 × 10–8

4.8 × 10–2

Inh Type M

Lung

4.1 × 10–8

1.5 × 10–1

2.3 × 10–8

8.5 × 10–2

Inh Type S

Lung

4.9 × 10–8

1.8 × 10–1

2.9 × 10–8

1.1 × 10–1

Inh Type F

Red marrow

1.5 × 10–9

5.6 × 10–3

4.3 × 10–9

1.6 × 10–2

Ing soluble

Red marrow

3.0 × 10–9

1.1 × 10–2

8.9 × 10–9

3.3 × 10–2

Inh Type M

Lung

1.5 × 10–8

5.6 × 10–2

7.1 × 10–9

2.6 × 10–2

Inh Type S

Lung

1.7 × 10–8

6.2 × 10–2

1.7 × 10–8

6.3 × 10–2

144

Ce, D

137Cs,

60

D

Co

285 d

30.2 y

5.27 y

βγ

βγ

βγ

NS, BC U

NS, BC, U

NS, BC, U

244

Cm

18.1 y

αγn

NS, IVC, U

Inh Type M

Lung

7.3 × 10–7

2.7 × 100

1.7 × 10–5

6.3 × 101

154

Eu

8.59 y

β, EC

NS, IVC, U, F

Inh Type M

Lung

2.5 × 10–8

9.2 × 10–2

3.2 × 10–8

1.2 × 10–1

Inh Type S

Lung

2.9 × 10–8

1.1 × 10–1

2.4 × 10–8

8.8 × 10–2

U

Inh or ing HTO

Red marrow

1.5 × 10–11

5.6 × 10–5

1.8 × 10–11

6.7 × 10–5

IVC, U

Inh vapor or ing

Thyroid

4.1 × 10–7

1.5 × 100

2.0 × 10–8

7.4 × 10–2

3

H

131

I, D

12.3 y

β

8.02 d

βγ

154 / 10. STAGE 5: INTERNAL CONTAMINATION ASSESSMENT

TABLE 10.4—Dose estimates for radionuclide intakes in adults (Section 20).

192Ir 103 32

Pd

P

238

Pu

73.8 d

βγ

17 d

βγ

14.3 d

β

87.7 y

239Pu

24,110 y

αn

α

NS, BC, U, F U IVC, U

IVC, U, F

IVC, U, F

1.7 × 10–8

6.3 × 10–2

4.1 × 10–9

1.5 × 10–2

Red marrow

3.3 × 10

–11

1.2 × 10

–4

2.9 × 10

1.1 × 10–3

Inh Type M

Lung

1.4 × 10

–8

5.2 × 10

–2

–9

2.9 × 10

1.1 × 10–2

Ing soluble

Red marrow

6.4 × 10–9

2.4 × 10–2

2.4 × 10–9

8.9 × 10–3

Lung

6.3 × 10

–7

2.3 × 10

0

3.1 × 10

1.1 × 102

Inh Type S

Lung

7.4 × 10

–7

2.7 × 10

0

–5

1.1 × 10

4.1 × 101

Inh Type M

Lung

5.4 × 10–7

2.0 × 100

3.3 × 10–5

1.2 × 102

Lung

6.4 × 10

–7

2.3 × 10

0

8.4 × 10

3.1 × 101

Lung

5.5 × 10

–7

2.0 × 10

0

–6

2.3 × 10

8.5 × 100

Red marrow

3.2 × 10–9

1.2 × 10–2

Kidneys

3.5 × 10–8

1.3 × 10–1

Red marrow

4.8 × 10–9

1.8 × 10–2

3.7 × 10–7

1.4 × 100

Kidneys

3.9 ×

1.4 ×

Lung

IV injection

Inh Type M

Inh Type S 210

Po

138 d

α

U, F

Inh Type M

Ing inorganic

226 188

103

Ra, D Re

Ru

1,600 y

αβγ

17 h

βγ

39.3 d

βγ

NS, IVC, U, F BC, U

NS, BC, U, F

D

374 d

βγ

NS, BC, U, F

IV injection

Inh Type M Inh Type S Inh Type M

–5

–6

10–1

4.7 × 10

–7

1.7 × 100

2.2 × 10–6

8.1 × 100

Stomach

5.3 × 10

–9

2.0 × 10

–9

1.4 × 10

5.2 × 10–3

Red marrow

1.1 ×

Lung

10–10

5.2 ×

–2

10–4

8.9 × 10

–9

3.3 × 10–2

1.8 × 10–9

6.6 × 10–3

Lung

1.0 × 10

–8

3.7 × 10

–9

2.1 × 10

7.7 × 10–3

Lung

3.3 ×

1.7 ×

10–8

6.3 × 10–2

10–8

1.2 ×

–2

10–1

/ 155

NS, BC, U, F 106Ru,

Inh Type M

10–8

–10

10.4 INTAKE AND DOSE ASSESSMENT

Lung

Inh Type M

Absorbed Dose (30 d) Radionuclidea

153Sm

90

Tc

232Th,

234

90

46.5 h

Sr, D

99m

28.8 y

6.02 h

D

Uf, D

Y

Half-Lifeb

1.4 × 1010 y

2.5 × 105 y

64.1 h

Main Radiationsc

βγ

β

βγ

αβγ

αβγ

β

Method of Measurementd

NS, BC, U

U, F

BC,

NS, IVC, U, F

NS, IVC, U, F

U

Exposuree Tissues

Effective Dose (50 y)

(Gy Bq–1)

(rad μCi–1)

(Sv Bq–1)

(rem μCi–1)

Inh Type S

Lung

3.9 × 10–8

1.4 × 10–1

3.4 × 10–8

1.3 × 10–1

Inh Type M

Lung

3.2 × 10–9

1.2 × 10–2

6.8 × 10–10

2.5 × 10–3

IV injection

Red marrow

5.8 × 10–10

2.1 × 10–3

2.9 × 10–10

1.1 × 10–3

Inh Type F

Red marrow

4.6 × 10–9

1.7 × 10–2

3.0 × 10–8

1.1 × 10–1

Ing soluble

Red marrow

4.0 × 10–9

1.5 × 10–2

2.8 × 10–8

1.0 × 10–1

IV injection

Stomach

6.2 × 10–11

2.3 × 10–4

1.9 × 10–11

7.0 × 10–5

Red marrow

4.2 × 10–12

1.6 × 10–5

Inh Type M

Lung

3.1 × 10–7

1.1 × 100

2.9 × 10–5

1.1 × 102

Inh Type S

Lung

3.7 × 10–7

1.4 × 100

1.2 × 10–5

4.4 × 101

Inh Type M

Lung

4.6 × 10–7

1.7 × 100

2.1 × 10–6

7.8 × 100

Inh Type S

Lung

5.4 × 10–7

2.0 × 100

6.8 × 10–6

2.5 × 101

Inh Type M

Lung

4.5 × 10–9

1.7 × 10–2

1.6 × 10–9

5.9 × 10–3

a Radionuclides are listed alphabetically by element. The letter D refers to the possible presence of daughters with a half-life of <25 y. The radiations of the daughters are not included in the listing. bRadioactive half-life.

156 / 10. STAGE 5: INTERNAL CONTAMINATION ASSESSMENT

TABLE 10.4—(continued)

cThe

10.4 INTAKE AND DOSE ASSESSMENT

primary radiations are listed. These include radiations emitted by dosimetrically-significant chain members. β = both positron and electron emission. γ = conversion x-ray emissions as well as gamma rays. EC = electron conversion n = neutrons d The following symbols are used to indicate principal techniques for measuring external contamination or indicating internal exposure. The order of the symbols has no significance in the listing: BC = whole-body count (standard gamma detection methods) F = feces sample analyses IVC = special in vivo counting techniques useful for low-energy counting (wound monitoring, thyroid counting), or special low-energy x-ray or gamma detectors for chest counts (e.g., plutonium or americium counting) NS = nasal swab counted in laboratory if inhalation suspected U = urine sample analyses e Inh = inhalation Ing = ingestion IV = intravenous HTO = tritiated water. Types F (fast), M (moderate), and S (slow) are absorption types of particulate aerosols in the respiratory tract, as defined in ICRP Publication 66 (ICRP, 1994a). f Uranium always comes as a mixture of the isotopes 238, 234 and 235. Natural uranium is composed of 99.3 % 238U, 0.711 % 235U, and 0.0058 % 234U by weight. In equilibrium, natural uranium contains the same activity of 238U and 234U (48.9 % each) and 235U (2.2 %). Enriched uranium is obtained when the concentration of 235U is increased to significantly >0.711 % by weight. When the concentration of 235U is decreased to significantly <0.711 % by weight (~0.2 to 0.3 % by weight), the material is called depleted uranium (Section 20.24).

/ 157

11. Stage 6: Clinical Decision Guidance (hospital)

Objectives • evaluate radiation doses with respect to Clinical Decision Guide (CDG); and • provide guidance to physicians making treatment decisions. See Figure 3.1 for flow of persons through all nine stages in the management of radionuclide contamination. 11.1 Introduction Decisions to treat individuals who have internal depositions of radionuclides are not always obvious, even when the internal contamination is well characterized. This is serious when there are one or a few contaminated individuals. It becomes more so when there are potentially large numbers of persons with internal contamination. Many factors comprise a physician’s decision to treat internal contamination cases. These include age of the patient, number of patients, the patient’s health status, time after intake, and confounding injuries. Generally most important is the radiation dose and the associated health risks. This section offers an approach that may assist physicians in making treatment decisions, whether for one contaminated individual or many. This approach involves the use of an operational quantity, the “Clinical Decision Guide” (CDG), based on internal dose considerations that allow for more precision in making treatment decisions. 11.2 Clinical Decision Guides The Clinical Decision Guide (CDG), a new operational quantity, is defined here to provide a measure that physicians can use when 158

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considering the need for medical treatment for internally-deposited radionuclides or as a screening level indicating the need for a more detailed investigation of tissue-specific absorbed doses over different time periods. For radionuclides other than isotopes of iodine, the CDG is the maximum, once-in-a lifetime intake of a radionuclide that represents: • stochastic risk, as judged by the calculated effective dose over 50 y for intake by adults and to age 70 y for intake by children, that is in the range of risks associated with guidance on dose limits for emergency situations (DOE, 2008a; FEMA, 2008; ICRP, 1991a; NCRP, 1993; 2005a) and • avoidance of deterministic effects as judged by the calculated 30 d RBE-weighted absorbed doses to red marrow and lungs, with allowance for the significant uncertainties often involved in an initial evaluation of the chemical and physical form of a radionuclide and the level of activity taken into the body during an incident. CDGs for radioiodine are defined differently from those for other radionuclides because the cumulative dose to the thyroid is the pertinent measure of risk in this case, and FDA (2001) has issued specific guidance regarding projected thyroid doses at which treatment for intake of radioiodine is indicated for different risk groups (Section 12). The Guidebook for the Treatment of Accidental Internal Radionuclide Contamination of Workers, a joint publication for CDC and DOE, recommended that treatment be considered for internal radionuclide depositions that would result in effective doses in the range of 20 to 200 mSv (2 to 20 rem) or higher (Bhattacharyya et al., 1992). ICRP has judged that no tissues are expected to express clinically-relevant functional impairment from internallydeposited radionuclides at absorbed doses up to ~100 mGy (10 rad), low or high linear energy transfer (LET), with threshold doses for deterministic effects in most tissues in the 1 Gy (100 rad) or higher range (ICRP, 2007). Based upon the recommendations and limits for emergency situations and knowledge of deterministic effects, the numerical values of dose used as a basis in this Report for computing the CDG intake values for different radionuclides, excluding isotopes of iodine, in adults are 0.25 Sv (25 rem) (50 y effective dose) for consideration of stochastic effects [based on the population-averaged nominal cancer fatality risk coefficient of 5 % Sv–1 derived from epidemiological data (ICRP, 2007), this represents about a 1.3 %

160 / 11. STAGE 6: CLINICAL DECISION GUIDANCE (HOSPITAL) lifetime risk of fatal cancer attributable to the radiation dose]; a 30 d RBE-weighted absorbed-dose value of 0.25 Gy-Eq (25 rad-Eq) for consideration of deterministic effects to bone marrow; and a 30 d RBE-weighted absorbed-dose value of 1 Gy-Eq (100 rad-Eq) for consideration of deterministic effects to the lungs. Thus, for intake of a radionuclide other than an isotope of iodine, the CDG for an adult is the maximum intake satisfying these dose constraints for both stochastic and deterministic effects: CDG = 0.25 Gy-Eq 1.0 0.25 Sv - , --------------------------------------------, ----------------------------------------------------------MIN ----------------------------–1 –1 –1 e ( Sv Bq ) d Red Marrow ( Gy-Eq Bq ) d Lung ( Gy-Eq Bq )

(11.1)

where: e = effective dose coefficient for the radionuclide dRed Marrow and dLung = RBE-weighted absorbed-dose coefficients for red marrow and lung, respectively MIN = minimum value of the three arguments The CDG for an adult is the intake that satisfies the constraint on the effective dose and the 30 d absorbed dose to the red marrow and lungs. For radionuclides other than isotopes of iodine, the CDGs for children (age 0 to 18 y) and pregnant women are defined as onefifth the adult value, reflecting the increased vulnerabilities during development and maturation (AAP, 2003). Children weighing >70 kg should be considered as adults. For intake or expected intake of radioiodine, FDA (2001) recommends that KI be administered to adults >40 y of age if the projected dose to thyroid is ≥5 Gy (500 rad), to adults 18 to 40 y of age if the projected dose is ≥0.1 Gy (10 rad), and to pregnant or lactating women or persons <18 y of age if the projected dose is ≥0.05 Gy (5 rad). In this Report, CDGs for 131I (the only isotope of iodine considered here) are derived separately for the following subgroups of the population, considering not only the FDA dose guidelines for different risk groups but also projected differences with age in dose per unit intake of radioiodine (Section 20): adults of age >40 y; adults 18 to 40 y; pregnant or lactating women; and age groups 12 to 18, 7 to 12, 3 to 7, 0.5 to 3, and <0.5 y. The dose coefficient for thyroid (committed equivalent dose to thyroid per unit intake) for a reference adult is applied to each of the first three subgroups, and the coefficients for intake ages 15 y, 10 y, 5 y, 1 y, and 3 months are applied to ages groups 12 to 18, 7 to 12, 3 to 7, 0.5 to 3, and <0.5 y,

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/ 161

respectively. The CDG for radioiodine for a specific subgroup of the population is defined as the FDA dose guideline value applicable to that subgroup, divided by the thyroid dose coefficient for that subgroup (Table 20.56). With the following exceptions, CDGs for radionuclides addressed in this Report are determined by the estimated risk of stochastic effects. CDGs for 103Ru (T1/2 = = 39.3 d) and 192Ir (T1/2 = 73.8 d) inhaled in either a Type M or S form and 153Sm (T1/2 = 46.5 h) or 32 P (T1/2 = 14.3 d) inhaled in a Type M form are determined by the 30 d RBE-weighted absorbed dose to the lungs. The CDGs for 32P ingested in a soluble form and 153Sm injected intravenously in a soluble form are determined by the 30 d RBE-weighted absorbed dose to the red bone marrow. For each of these cases a major portion of the committed dose to lungs, red marrow, and other tissues is delivered over the first 30 d after intake. CDGs are intended to serve as one measure of the possible need for early treatment of individuals with elevated intake of a radionuclide, especially in mass casualty situations. CDGs are set at cautiously low levels, particularly in relation to threshold values for deterministic effects. This is based on the consideration that initial evaluation of the chemical and physical form of the radionuclide and the level of activity taken into the body following exposure to radionuclides often involves significant errors, as revealed by more detailed follow-up measurements. For radiation dose estimates of the magnitude indicated in the definition of CDG, the increase in risk of stochastic effects and conceivably even deterministic effects may become an important consideration with regard to a clinical course of action. CDG-based clinical actions will vary, however, depending on available time and resources (as determined, for example, by the number of exposed persons and the extent of life-threatening injuries). A physician may choose to use 1 CDG as a basis for treatment or as a screening level indicating the need for a more detailed investigation of tissuespecific absorbed doses over different time periods. In summary, with respect to the patient, factors to be considered include trauma that might have occurred as a result of the incident, state of general health, age, pregnancy, emotional state, the route of intake, the time since intake, and the biochemical and physical properties of the internally-contaminating radionuclide and its site of deposition in the body. With respect to the incident, the number of people contaminated is an important consideration because of logistics and the availability of resources including medication. Physicians may depend on chest or whole-body counts, or on measurements of activity in excreta or on nasal swabs to determine

162 / 11. STAGE 6: CLINICAL DECISION GUIDANCE (HOSPITAL) the level of intake to which a patient has been exposed. Tables 11.1 and 11.2 give levels of measured activity in urine, chest, total body, or on nasal swabs that are indicative of an intake of 1 CDG for each of the radionuclides and routes of contamination addressed in Section 20. It is beyond the scope of this Report to provide a comprehensive tabulation of CDGs and measurable indicators of 1 CDG for all radionuclides of potential concern. The radionuclides addressed in Tables 11.1 and 11.2 are those considered most likely to be present in significant quantities in a contamination incident, based on past experience as well as current uses, abundance and availability. The specific forms (e.g., tritiated water) or generic solubility levels (e.g., Type F, M or S) considered for each radionuclide also are those considered most likely to be encountered, insofar as such information is available. Inhalation of Type M material is considered for several radionuclides in lieu of specific information on likely airborne forms because Type M generally provides more robust estimates of tissue dose than do the other standard absorption types. Estimates for inhalation of Type S material also are included for some radionuclides that have been encountered in relatively-insoluble form in occupational settings. The size of an inhaled particulate aerosol typically cannot be characterized with much certainty during the early phase following a radiological incident. A particle size of 5 µm AMAD was assumed in the derivation of CDGs for inhaled particulate aerosols because this is the ICRP default value for occupational intakes and it is consistent with extensive occupational experience (Dorrian and Bailey, 1995; ICRP, 2002a). It is also judged to be a reasonable default particle size for radiological incidents involving members of the public because intakes of clinical significance are more likely to occur near the point of release where the sizes of airborne particles are usually the largest and the concentrations of airborne radionuclides are usually highest. At distances farther from the release site, sedimentation and other processes will have removed a large fraction of the larger particles and the diffusion of the plume will have decreased radionuclide concentration. CDGs and associated excretion and retention values for intake of selected forms and modes of intake of 60Co, 90Sr, 137Cs, 192Ir, 238Pu, 241Am, and 252Cf can be calculated as described in Example 3, Method 2 below for 1 µm AMAD particles using model predictions for members of the public in Section 20. As illustrated in Table 11.3, the CDG and retention and excretion values indicative of 1 CDG for an inhaled radionuclide generally are not strongly sensitive to the inhaled particle size in the range 0.5 to 10 µm AMAD.

TABLE 11.1—Model predictions used to assess whether a radionuclide intake exceeds the CDG.a (Values are for a reference adult. For application to children and pregnant women, the CDG and the activities in urine, chest, total body, and nasal passages that correspond to 1 CDG should be divided by five). CDG (intake activity)

Effective Dose Coefficient Intake Modeb

Form

(Sv Bq–1)

3 3

H

Inhalation

H

Ingestion

Early Excretion and Retention (percentage of intake)

(mrem μCi–1)

(Bq)

Urinary Excretion 0 – 24 h

(μCi)

Total-Body Retention at 24 h

Urinary Excretion 0 – 24 h

NAg

94

3.0 × 1010

94

3.0 × 10

10 8

Retention in Chest at 24 he

Total-Body Retention at 24 h

Nasal Swab Soon After Inhalationf

NA

7.8 × 1011

NA

NA

7.8 × 10

11

NA

2.4 × 10

2.2 × 10

9

1.9 × 109

NA

1.0 × 10

9

1.1 × 108

5.6 × 107 4.3 × 108

4.4 × 107

6.7 × 10–2

1.4 × 1010 3.8 × 105

HTO

1.8 × 10

6.7 × 10

–2

1.4 × 10

3.8 × 10

5

1h

7.1 × 10

1.9 × 10

3

3.7

5.5

51

1.6 × 10

8.7

NA

80

2.0 × 108

NA

5.8

49

4.2 × 10

1.2 × 10

6.4

49

–11

10

Type M

1.4 × 10

5.2 × 10

32

Ingestion

Soluble

6.4 × 10–9 i

2.4 × 101 i

3.9 × 107

1.1 × 103

3.5 × 10

9.5 × 10

–8 h

7

3.6 3.6

NA

8

2.1 × 108

60

Inhalation

Type M

7.1 × 10

2.6 × 10

60

Inhalation

Type S

1.7 × 10–8

6.3 × 101

1.5 × 107

4.0 × 102

90

Inhalation

Type F

3.0 × 10–8

1.1 × 102

8.3 × 106

2.3 × 102

6.8

NA

49

3.4 × 107

NA

2.5 × 108

2.5 × 107

90

Ingestion

Soluble

2.8 × 10–8

1.0 × 102

8.9 × 106

2.4 × 102

5.6

NA

73

3.0 × 107

NA

3.9 × 108

NA

90

Inhalation

Type M

1.6 × 10–9

5.9 × 100

1.6 × 108

4.2 × 103

0.30

4.4

38

2.8 × 107

NA

78

7.2 × 109

5.7

49

3.6 × 107

Co Co Sr Sr Y

–9

1

7

2

103

Pd

IV injection Soluble

2.9 × 10–10

1.1 × 100

8.6 × 108

2.3 × 104

103

Ru

Inhalation

Type M

8.9 × 10–9 h

3.3 × 101 h 1.1 × 108

3.0 × 103

Ru

Inhalation

Type S

1.0 × 10

3.7 × 10

106

Ru

Inhalation

Type M

1.7 × 10–8

106

Ru

Inhalation

Type S

3.4 × 10–8

–8 h

—j

14 0.54

1.0 × 10

2.7 × 10

3

—

6.3 × 101

1.5 × 107

4.0 × 102

0.54

1.3 × 102

7.4 × 106

1.9 × 102

1h

8

j

—j

—j

j

6.3

48

—

5.8

49

4.8 × 106

6.4

49

—j

8

4.1 × 108 3.5 × 109 4.0 × 1010

NA

4.7 × 108 NA

3.8 × 108 3.2 × 109

3.4 × 108

3.8 × 10

9

3.0 × 108

5.1 × 107 4.3 × 108

4.4 × 107

2.8 × 107 2.2 × 108

2.2 × 107

8

2.9 × 10

/ 163

103

2.0

7

11.2 CLINICAL DECISION GUIDES

1.8 × 10–11

Inhalation

P

Retention in Chest at 24 he

HTO

32

P

Early Excretion and Retention Levels Indicative of Intake of 1 CDG (dpm)d

b,c

CDG (intake activity)

Effective Dose Coefficient Intake Modeb

Early Excretion and Retention (percentage of intake)

Formb,c (Sv Bq–1)

(mrem μCi–1)

(Bq)

Urinary Excretion 0 – 24 h

(μCi)

Retention in Chest at 24 he

TotalBody Retention at 24 h

Early Excretion and Retention Levels Indicative of Intake of 1 CDG (dpm)

Urinary Excretion 0 – 24 h

Retention in Chest at 24 he

TotalBody Retention at 24 h

Nasal Swab Soon After Inhalationf

137

Cs

Inhalation

Type F

4.3 × 10–9

1.6 × 101

5.8 × 107

1.6 × 103

2.2

NA

58

7.7 × 107

NA

2.0 × 109

1.7 × 108

137

Cs

Ingestion

Soluble

8.9 × 10–9

3.3 × 101

2.8 × 107

7.6 × 102

4.5

NA

95

7.6 × 107

NA

1.6 × 109

NA

1.1 × 10

2.9 × 10

3.7 × 10

3.2 × 10

8

3.2 × 107

Ce

Inhalation

Type M

2.3 × 10

8.5 × 10

144 Ce

Inhalation

Type S

2.9 × 10–8

1.1 × 102

8.6 × 106

153

Inhalation

Type M

3.2 × 10–9 h

1.2 × 101 h 3.1 × 108

144

Sm

–8

1

7

0.052

5.7

50

3.4 × 10

2.3 × 102

0.00088

6.4

49

4.6 × 103

3.3 × 107 2.5 × 108

2.6 × 107

8.5 × 103

0.16

4.0

35

3.0 × 107

7.5 × 108 6.6 × 109

9.4 × 108

2

5

7

153Sm

IV injection Soluble

5.8 × 10–10 i 2.2 × 100 i

4.3 × 108

1.2 × 104

5.8

NA

65

1.5 × 109

154

Eu

Inhalation

Type M

3.2 × 10–8

1.2 × 102

7.8 × 106

2.1 × 102

0.50

5.8

49

2.3 × 106

2.7 × 107 2.3 × 108

2.3 × 107

154Eu

Inhalation

Type S

2.4 × 10–8

8.9 × 101

1.0 × 107

2.8 × 102

0.0084

6.4

49

5.2 × 104

4.0 × 107 3.1 × 108

3.1 × 107

188

IV injection Soluble

1.4 × 10–9

5.2 × 100

1.8 × 108

4.8 × 103

9.0

NA

31

9.6 × 108

Inhalation

Type M

1.7 × 10–8 h

6.3 × 101 h 5.9 × 107

1.6 × 103

0.31

5.7

49

1.1 × 107

2.0 × 108 1.7 × 109

1.8 × 108

Inhalation

Type S

2.0 × 10

7.4 × 10

1.9 × 10

1.5 × 10

1.5 × 108

Ingestion

Soluble

3.7 × 10–7

3.0 × 107

NA

3.2 × 10

6

3.3 × 105

4.0 × 105 3.4 × 106

3.4 × 105

Re

192Ir 192

Ir

210Po

–8 h

5.0 × 10

1.4 × 10

0.038

6.4

49

1.1 × 10

1.4 × 103

6.8 × 105

1.8 × 101

0.059

NA

73

2.4 × 104

NA 3.7 × 10

Inhalation

Type M

2.3 × 10

8.5 × 10

226Ra

Inhalation

Type M

2.2 × 10–6

8.1 × 103

–6

3.3 × 109

NA

3

1h

Po

210

1.7 × 1010

NA

3

7

6

1.1 × 10

2.9 × 10

0

0.12

5.7

50

7.8 × 10

1.1 × 105

3.1 × 100

0.16

5.8

50

1.1 × 104

5

3

8

5

9

NA

NA

164 / 11. STAGE 6: CLINICAL DECISION GUIDANCE (HOSPITAL)

TABLE 11.1—(continued)

Inhalation

Type M

4.8 × 10–5

1.8 × 105

5.2 × 103

1.4 × 10–1

0.18

5.8

50

5.6 × 102

1.8 × 104 1.6 × 105

1.6 × 104

232

Th

Inhalation

Type M

2.9 × 10–5

1.1 × 105

8.6 × 103

2.3 × 10–1

0.11

5.8

50

5.7 × 102

3.0 × 104 2.6 × 105

2.6 × 104

232Th

Inhalation

Type S

1.2 × 10–5

4.4 × 104

2.1 × 104

5.6 × 10–1

0.0013

6.4

49

1.6 × 101

8.0 × 104 6.1 × 105

6.3 × 104

238

Uk

Inhalation

Type M

2.1 × 10–6

7.8 × 103

1.2 × 105

3.2 × 100

2.3

5.8

48

1.6 × 105

4.1 × 105 3.4 × 106

3.6 × 105

238U k

Inhalation

Type S

6.8 × 10–6

2.5 × 104

3.7 × 104

9.9 × 10–1

0.07 m

6.4

49

1.5 × 103 m 1.4 × 105 1.1 × 106

1.1 × 105

238

Pu

Inhalation

Type M

3.1 × 10–5

1.1 × 105

8.1 × 103

2.2 × 10–1

0.021

5.80

50

1.0 × 102

2.8 × 104 2.4 × 105

2.4 × 104

238Pu

Inhalation

Type S

1.1 × 10–5

4.1 × 104

2.3 × 104

6.1 × 10–1

0.00021 m

6.4

49

2.9 × 100 m 8.7 × 104 6.7 × 105

6.8 × 104

239

Pu

Inhalation

Type M

3.3 × 10–5

1.2 × 105

7.6 × 103

2.0 × 10–1

0.021

5.8

50

9.6 × 101

2.6 × 104 2.3 × 105

2.3 × 104

239Pu

Inhalation

Type S

8.4 × 10–6

3.1 × 104

3.0 × 104

8.0 × 10–1

0.00021 m

6.4

49

3.8 × 100 m 1.1 × 105 8.8 × 105

8.9 × 104

Type M

2.7 × 10–5

1.0 × 105

9.3 × 103

2.5 × 10–1

0.18

5.8

50

1.0 × 103

3.2 × 104 2.8 × 105

2.8 × 104

241

Am Inhalation

244Cm

Inhalation

Type M

1.7 × 10–5

6.3 × 104

1.5 × 104

4.0 × 10–1

0.18

5.8

50

1.6 × 103

5.1 × 104 4.4 × 105

4.4 × 104

252

Inhalation

Type M

1.1 × 10–5

4.1 × 104

2.3 × 104

6.1 × 10–1

0.27

5.8

50

3.7 × 103

7.9 × 104 6.8 × 105

6.8 × 104

Cf

/ 165

a For radionuclides other than isotopes of iodine, CDG for a specific form of a radionuclide and mode of exposure is the intake activity estimated to result in the most restrictive of the following doses to an adult: a 50 y effective dose of 0.25 Sv (25 rem), an RBE-weighted 30 d absorbed dose to red marrow of 0.25 Gy-Eq (25 rad-Eq), or an RBE-weighted 30 d absorbed dose to lung of 1 Gy-Eq (100 rad-Eq). Fivefold lower CDGs are applied to children and pregnant women. A radiation weighting factor of 20 is applied to the alpha absorbed dose in deriving the effective dose coefficient and RBE values of two and seven are applied in deriving the 30 d RBE-weighted absorbed dose values for red marrow and lungs, respectively. Effective dose is more restrictive than the 30 d RBE-weighted absorbed dose to red marrow or lung in most cases. b HTO is tritiated water. Types F, M, and S refer to “absorption types” of inhaled material as defined by ICRP (1994a) and represent material that dissolves at fast, moderate, or slow rates in the respiratory tract and is absorbed to blood at relatively-high, moderate, or relatively-low levels, respectively. For inhaled particulate material, the particle-size distribution is assumed to be lognormal with a 5 μm AMAD. The term IV injection refers to intravenous injection. c If no information is available regarding the mode of intake or form of the radioactive material taken into the body, and if multiple cases are provided in this table for the radionuclide of concern, then measurement of activity in urine, chest, total body, or nasal swipe should be compared with the smallest listed value for urine, chest, total body, or nasal swipe, respectively, given for that radionuclide. d Divide by 60 to convert to becquerels and by 2.22 × 106 to convert to microcuries.

11.2 CLINICAL DECISION GUIDES

227Ac

e Retention in the chest refers to activity measured externally over the thoracic portion of the respiratory tract. This is assumed to represent activity in the lungs. fThe portion of inhaled activity found in a nasal swab in the early hours after inhalation of a radionuclide is highly variable, depending on such factors as aerosol size, the extent of nose breathing versus mouth breathing during the exposure, and the amount of nose blowing and wiping of the nostrils since the beginning of exposure. The listed activity for nasal swab represents 5 % of the inhaled amount, based upon experimental data and model predictions summarized in Section 10.3.11 and in greater detail in Section 16.4.1.2. The presence of radioactivity in a nasal swab is suggestive evidence of an inhalation exposure, particularly if both nares are contaminated. The absence of activity in a nasal swab does not establish that there was no inhalation. g NA = not applicable. In many cases of inhalation of radionuclides, external counts over the chest are not useful. For example, after inhalation of highly-soluble forms of radionuclides, activity quickly moves from the lungs to blood. Also, radionuclides that emit little if any penetrating radiation (e.g., the beta-emitter 3H or the alpha-emitter 210Po) are not detectable by external measurement. h The indicated dose is the RBE-weighted 30 d absorbed dose to the lungs, which is more restrictive than the effective dose in this case. iThe indicated dose is the RBE-weighted 30 d absorbed dose to red marrow, which is more restrictive than the effective dose in this case. j For these cases, calculation of an intake based on urinary excretion data is not recommended because of the high sensitivity of the estimate to the GI absorption fraction, which is not well established. Where feasible, decisions concerning treatment should be based on external measurement of activity in the chest, supplemented with measurement of activity in feces. Fecal excretion data can be interpreted on the basis of tabulations in Section 20. k Table entries for 238U may also be applied to 234U or 235U. Chemical toxicity of uranium (nephrotoxicity) is generally of greater immediate concern than radiological toxicity following acute inhalation of elevated quantities of natural or depleted uranium but not necessarily for inhalation of enriched uranium. lThis letter intentionally not used in table to avoid confusion. m Measurement of the urinary excretion rate should be supplemented with measurement of fecal excretion rate where feasible. A number of inhalation cases have been reported in which little or no activity was measured in urine for an extended period following significant exposure to an insoluble form of this radionuclide (see case studies for uranium and plutonium in Section 20).

166 / 11. STAGE 6: CLINICAL DECISION GUIDANCE (HOSPITAL)

TABLE 11.1—(continued)

TABLE 11.2—Model predictions used to assess whether an intake of 131I by inhalation as a vapor or ingestion in a soluble form exceeds CDG.a Committed Equivalent Dose to Thyroid

CDG (intake activity)

Excretion and Retention During First 24 h (percentage of intake)

Excretion and Retention Levels During First 24 h Indicative of an Intake of 1 CDG (dpm)b

Group (mrem μCi–1)

(Bq)

(μCi)

Urinary Excretion 0 – 24 h

Retention in Thyroid at 24 h

Total-Body Retention at 24 h

Urinary Excretion 0 – 24 h

Retention in Thyroid at 24 h

Total-Body Retention at 24 h

Adult >40 y

3.9 × 10–7

1.4 × 103

1.3 × 107

3.5 × 102

56

23

33

4.3 × 108

1.8 × 108

2.5 × 108

Adult 18 – 40 y

3.9 × 10–7

1.4 × 103

2.6 × 105

6.9 × 100

56

23

33

8.6 × 106

3.5 × 106

5.1 × 106

Pregnancy or lactation

3.9 × 10–7

1.4 × 103

1.3 × 105

3.5 × 100

56

23

33

4.3 × 106

1.8 × 106

2.5 × 106

Age 12 – 18 y

6.2 × 10–7

2.3 × 103

8.1 × 104

2.2 × 100

56

23

33

2.7 × 106

1.1 × 106

1.6 × 106

Age 7 – 12 y

9.5 × 10–7

3.5 × 103

5.3 × 104

1.4 × 100

56

23

33

1.8 × 106

7.3 × 105

1.0 × 106

Age 3 – 7 y

1.9 × 10–6

7.0 × 103

2.6 × 104

7.1 × 10–1

56

23

33

8.8 × 105

3.6 × 105

5.2 × 105

Age 0.5 – 3 y

3.2 × 10–6

1.2 × 104

1.6 × 104

4.2 × 10–1

56

22

32

5.3 × 105

2.1 × 105

3.0 × 105

Age <0.5 y

3.3 × 10–6

1.2 × 104

1.5 × 104

4.1 × 10–1

56

22

29

5.1 × 105

2.0 × 105

2.6 × 105

aThe tabulated values are based on threshold doses estimated by FDA (2001) and listed in Table 12.14 for different risk groups, together with age-specific biokinetic and dose estimates for 131I inhaled as a vapor listed in Table 20.56 of this Report. b Divide by 60 to convert to becquerels and by 2.22 × 106 to convert to microcuries.

11.2 CLINICAL DECISION GUIDES

(Sv Bq–1)

/ 167

Range of Values Expressed as Multiple of Value for 5 µm AMADa Excretion and Retention Levels During First 24 h Indicative of Intake of 1 CDG

Radionuclide Form

CDG Urinary Excretion 0 – 24 h

Retention in Chest at 24 h

Total-Body Retention at 24 h

Sr

Type F

0.97 – 1.5

0.99 – 1.0

NAb

0.71 – 1.1

137Cs

Type F

1.0 – 1.9

0.98 – 1.0

NA

0.90 – 1.0

238

Type S

0.64 – 1.8

0.46 – 1.4

0.82 – 1.4

0.34 – 1.7

Type M

0.64 – 1.6

0.46 – 1.3

0.71 – 1.3

0.33 – 1.4

90

Pu

241Am

example, for the CDG for the range 1.0 to 1.9 indicates that the calculated CDGs range from 1.0 × 5.8 × 107 to 1.9 × 5.8 × 107 Bq for particle sizes in the range 0.5 to 10 µm AMAD, where 5.8 × 107 Bq is the CDG for inhaled 137Cs, Type F, 5 µm AMAD. b NA = not applicable to this case due to rapid removal of activity from lungs. aFor

137Cs,

168 / 11. STAGE 6: CLINICAL DECISION GUIDANCE (HOSPITAL)

TABLE 11.3—Illustrations of sensitivity of CDG and measurable indicators of 1 CDG to size of inhaled particulate aerosols in the range 0.5 to 10 µm AMAD.

11.4 EXTRAPOLATION OF DATA

/ 169

11.3 Clinical Decision Guide Instrument Considerations Radiation detection instruments generally readout in counts per minute. To relate these readings to the quantity of radioactivity in a given sample expressed in disintegrations per minute, the instruments must be calibrated. This is accomplished by reference to the counting efficiency, the number of counts per minute corresponding to a given quantity of activity in disintegrations per minute, determined for each instrument under an applicable geometry. For example, if a detector has a 40 % counting efficiency and the CDG for a given radionuclide corresponds to 1,000 dpm, then 1 CDG (i.e., 1,000 dpm) will produce an instrument count rate of 400 cpm. Accordingly, for this radionuclide, a reading of 400 cpm on a radiation detector will reflect an uptake of 1 CDG. Some radionuclides (e.g., 3H as tritiated water and 137Cs in soluble form) are quickly absorbed and distributed throughout the body. Accordingly, these radionuclides will not remain in the lungs, even if inhaled, and the results from chest count measurements will not accurately reflect the amount of inhaled activity. In such cases, it is more appropriate to collect urine samples or to conduct whole-body counting. Radionuclides in the body that are primarily beta and alpha emitters (e.g., 3H, 32P, 90Sr, 210Po) will be difficult to detect using conventional radiation detection instruments. 11.4 Extrapolation of Data for Spot Urine Sample to 24 h Values All excretion values in Tables 11.1 and 11.2 and other tables in this Report refer to 24 h samples. In responding to a mass casualty incident it may only be possible to obtain a spot urine sample. Ageand gender-specific reference values for 24 h urine volumes are listed in Table 11.4 (ICRP, 2002b). If a spot urine sample is obtained, the measured activity can be extrapolated to a 24 h value based on an age and gender matched reference 24 h urine volume. If the patient is hospitalized or if resources permit, a true 24 h urine volume may be obtained. A 24 h urine sample for healthphysics analysis is collected in the same manner as for medical analysis. The patient voids, the time is recorded, and the urine is then collected for the next 24 h. The collection ends with the patient collecting a final urine void 24 h after the start of the collection period. Use of this method in health-physics internal-dose calculations serves two purposes: it eliminates the need for a dynamic bladder model and it integrates out diurnal variation of excretion.

170 / 11. STAGE 6: CLINICAL DECISION GUIDANCE (HOSPITAL) TABLE 11.4—Reference values for 24 h urine volume (ICRP, 2002b). Excretion (mL d-1) Age Male

Female

Newborn

300

300

1y

400

400

5y

500

500

10 y

700

700

15 y

1,200

1,200

Adult

1,600

1,200

NCRP Report No. 87 (NCRP, 1987) provides several correction methods that may be used to standardize activity if the collected volume is not that of a 24 h sample: • In a mass casualty terrorism incident involving collection of large numbers of spot specimens, the simplest correction is a straightforward volume correction. This correction is the easiest and the least accurate, but probably satisfactory for triage in a mass casualty incident. For example, suppose an adult male patient provides a 100 mL urine sample containing 2 µCi activity. The estimated activity for use in the tables (which require a 24 h void) is therefore: 1,600/100 × 2 µCi = 32 µCi. • Another correction validated in many metabolic studies is to adjust the activity in the spot urine sample to the creatinine content expected for a true 24 h. Reference age- and genderspecific values for 24 h urinary excretion of creatinine are listed in Table 11.5. The corrected 24 h value is calculated as follows: CE corrected activity = ( observed activity in spot sample ) ⎛ --------------⎞ ⎝ C Obs ⎠

(11.2)

where: CE = expected creatinine for a 24 h sample CObs = observed creatinine value for the spot sample This method has the disadvantage of requiring an additional laboratory step (Boeniger et al., 1993; Graul and Stanley, 1982; Jackson, 1966; Kim, 1995).

11.5 WORKED EXAMPLES WITH BIOASSAY DATA

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TABLE 11.5—Reference values for 24 h urinary excretion of creatinine (ICRP, 2002b). Amount Excreted (g d-1) Age (y) Males

Females

Newborn

0.05

0.05

1

0.11

0.11

5

0.33

0.33

10

0.65

0.65

15

1.4

1.0

Adult

1.7

1.0

• A third correction method involves adjusting for the urine’s specific gravity. This is unwieldy in a mass casualty incident. This method is described in NCRP Report No. 87 (NCRP, 1987) and in the comprehensive monograph by Potter (2002). 11.5 Worked Examples with Bioassay Data Example 1: Inhalation of 60Co by an adult A man is admitted to an emergency unit a few hours after acute exposure to airborne 60Co, thought to be an oxide and presumably having a relatively-low level of solubility in the lungs. There are no medical or surgical issues, and he is externally decontaminated. Subsequent external measurements indicate that total-body activity is <106 dpm and activity in the chest is <5 × 105 dpm. Measurement of urinary 60Co indicate that 24 h excretion is <104 dpm. The total-body and chest measurements are considerably lower than the reference 24 h retention and excretion values in Table 11.1 corresponding to inhalation of 1 CDG of a relatively-insoluble form of 60Co (Type S). The urine measurement and total-body and chest measurements are also considerably lower than the reference values for inhalation of 60Co in moderately-soluble form (Type M). Thus, it appears that the patient has inhaled considerably <1 CDG and treatment would not likely be considered. Example 2: Inhalation of 137Cs by an adult A woman is admitted to the emergency department (ED) the first day after a 137Cs RDD incident involving a presumably soluble form of 137Cs. She is found to be contaminated from the waist up

172 / 11. STAGE 6: CLINICAL DECISION GUIDANCE (HOSPITAL) and to have some facial contamination, suggestive of an inhalation intake of 137Cs. There are no medical or surgical issues, and she is externally decontaminated by showering in the ED. The woman is then given equipment for a 24 h urine collection. The 24 h urine samples are counted the next day at a radiochemical laboratory and the sample is found to contain 1,850 kBq (50 μCi). The most direct use of Table 11.1 is to compare the 1,850 kBq (50 μCi) with the 24 h urinary excretion value associated with the CDG for Type F 137Cs in Column 11. For this comparison the excretion value in Table 11.1 in disintegrations per minute is converted to becquerel by dividing by 60 (1 Bq = 60 dpm): 7.7 × 107/60 = 1,283 kBq (35 μCi). Since the value of the 24 h urine sample is 1.44 times greater than the CDG related urinary excretion value in Table 11.1, the attending physician should consider treating the patient with Prussian blue. (Note: an uptake of this magnitude would likely be detected with a portable survey meter.) An equivalent approach is to compare the estimated uptake with the CDG value. From Table 11.1, the fraction excreted during day one is 0.022 for a Type F aerosol. The point estimate of intake for this woman is 1,850 kBq/0.022 = 8.4 × 107 Bq (50 μCi/0.0.022 = 2,270 μCi). Again this value is 1.44 times the CDG for an adult of 5.8 × 107 Bq (1,600 μCi) and treatment with Prussian blue should be considered. Example 3: Exposure of a child to an unknown form of 137Cs A 6 y old child is admitted to an emergency unit soon after acute exposure to airborne 137Cs of unknown particle size or chemical form. The child is externally decontaminated by scrubbing. External measurements at 24 h indicate total-body activity of about 2.5 × 108 dpm. The concentration of activity does not appear to be greater over the lungs than other parts of the body, suggesting that any inhaled 137Cs was in a soluble form. Urinary excretion of 137Cs during the first 24 h is estimated as 1.0 × 107 dpm. Two different methods are used to calculate the CDG and associated 24 h excretion and retention values. Method 1. The default approach is taken [i.e., the CDG and associated excretion and retention values for the child are taken as one-fifth the values given in Table 11.1 for inhalation of 137Cs (Type F, 5 μm AMAD) by a reference adult]: • CDG = 5.8 × 107 Bq / 5 = 1.2 × 107 Bq; • corresponding urinary 137Cs over 0 to 24 h = 7.7 × 107 dpm / 5 = 1.5 × 107 dpm; and

11.5 WORKED EXAMPLES WITH BIOASSAY DATA

• corresponding total-body 4.0 × 108 dpm.

137

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Cs at 24 h = 2.0 × 109 dpm / 5 =

The estimated 24 h urinary excretion rate and total-body retention at 24 h in the child (1.0 × 107 dpm and 2.5 × 108 dpm, respectively) are about two-thirds of the levels corresponding to 1 CDG. Method 2. It is assumed that the particle size is substantially smaller than 5 μm AMAD. A CDG and corresponding 24 h excretion and retention values are calculated from reference dosimetric and bioassay data for inhalation of 137Cs (Type F, 1 μm AMAD) by members of the public, given in Section 20, Table 20.30. When agespecific data are available, the CDG is still calculated for the reference adult and divided by five to find CDG for a child, but in this case biokinetic predictions for the child can be used to calculate the excretion and retention values corresponding to the derived CDG. For the reference adult, Table 20.30 gives an effective dose coefficient of 3.0 × 10–9 Sv Bq–1 and 30 d absorbed dose coefficients of 4.4 × 10–10 and 1.0 × 10–9 Gy Bq–1 for lungs and red marrow, respectively. These values are used in Equation 11.1 to calculate the CDG for the adult. The effective dose is found to be the limiting quantity: 7 0.25 Sv CDG for the adult = ------------------------------------------------ = 8.3 × 10 Bq . –9 –1 3.0 × 10 Sv Bq

(11.3)

The CDG for a child is one-fifth this amount, and the corresponding 24 h excretion and retention values for a 6 y old child can be estimated from reference biokinetic data given in Table 20.30 for 5 y of age. For a 5 y old child the predicted 24 h urinary excretion of 137Cs is 2.4 % of intake and whole-body retention at 24 h is 39 % of intake. Hence, for this age group: • CDG = 8.3 × 107 Bq / 5 = 1.7 × 107 Bq; • corresponding urinary 137Cs over 0 to 24 h = 0.024 × 1.7 × 107 Bq × 60 dpm/Bq = 2.4 × 107 dpm; and • corresponding total-body 137Cs at 24 h = 0.39 × 1.7 × 107 Bq × 60 dpm/Bq = 4.0 × 108 dpm. Total-body retention at 24 h corresponding to 1 CDG is the same as calculated by the first method. The 24 h urinary excretion value based on the second method is about 60 % higher than that based on the first method; hence, the first method gives the more cautious value of urinary 137Cs as a measure of the CDG in this case. Both methods indicate that the intake is slightly lower than the CDG

174 / 11. STAGE 6: CLINICAL DECISION GUIDANCE (HOSPITAL) and that treatment with Prussian blue might be considered but is not strongly suggested. The difference in the two derived values for urinary 137Cs is due almost entirely to the difference in predicted 137 Cs excretion rates in adults and 5 y old children. The derived values are relatively insensitive to the difference in the assumed particle size (5 μm in the first method and 1 μm in the second method). Example 4: Inhalation of 238Pu by a worker Plutonium-238 contamination is discovered in a laboratory where a chemist had been working with 238Pu nitrate 48 h earlier. A 24 h urine sample from the chemist representing 48 to 72 h postexposure shows a 238Pu content of 1.2 Bq (72 dpm). Occupational and experimental data for inhalation of plutonium nitrate generally are consistent with Type M behavior (Section 20.15). Table 11.1 gives a CDG of 8,100 Bq for inhalation of 238Pu of Type M. The 24 h urinary excretion rate indicative of 1 CDG is not given in Table 11.1 for the measurement period 48 to 72 h, but Section 20, Table 20.77 lists 24 h urinary excretion of 0.0068 % for day three following acute inhalation of 238Pu of Type M (5 μm AMAD) by a worker. Therefore, urinary 238Pu on day three indicative of 1 CDG for this case is 0.000068 × 8,100 Bq = 0.55 Bq (33 dpm). Thus, CDG appears to be exceeded by more than a factor of two based on the single urine measurement. Therefore, treatment with DTPA should be considered. Example 5: Mass exposure to airborne 137Cs Over 300 people including several children and a pregnant woman are potentially exposed to airborne 137Cs in an RDD incident in an indoor shopping mall. None of them receive cuts or have other immediate medical issues. During the first 24 h after the incident they are screened for signs of 137Cs contamination. Over 100 persons show potentially significant contamination levels. They are admitted to an emergency unit and decontaminated externally. A spot urine sample is collected from each person during one of the next 3 d (days two, three or four) and analyzed for 137Cs. The measured 137Cs activity in each sample is extrapolated to a 24 h excretion value based on the age and gender matched reference 24 h urine volume listed in Table 11.4. Solubility measurements made on radioactive material collected at the shopping mall indicate that the released 137Cs was in highly-soluble form (Type F). Table 11.1 lists a CDG of 5.8 × 107 Bq for inhalation of 137Cs of Type F by an adult. Table 20.27 of Section 20, lists predicted 24 h urinary 137Cs values of 0.71, 0.51, and 0.36 % of inhaled material on

11.5 WORKED EXAMPLES WITH BIOASSAY DATA

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days two, three, and five, respectively following acute inhalation of Cs of Type F (5 μm AMAD) by a reference adult. A urinary 137Cs value of 0.44 % for day four is estimated by averaging the values for days three and five. Therefore, for adults, 24 h urinary 137Cs corresponding to 1 CDG is:

137

• 0.0071 × 5.8 × 107 Bq = 4.1 × 105 Bq = 2.5 × 107 dpm for day two,; • 0.0051 × 5.8 × 107 Bq = 3.0 × 105 Bq = 1.8 × 107 dpm for day three; and • 0.0044 × 5.8 × 107 Bq = 2.6 × 105 Bq = 1.6 × 107 dpm for day four. For the children and pregnant women, the 24 h urinary 137Cs for day two, three or four corresponding to 1 CDG is one-fifth the value for the adult for that day: • 2.5 × 107 dpm / 5 = 5.0 × 106 for day two; • 1.8 × 107 dpm / 5 = 3.6 × 106 for day three; and • 1.6 × 107 dpm / 5 = 3.2 × 106 for day four.

12. Stage 7: Medical Management (hospital)

Objectives • determine and implement appropriate treatment; • evaluate treatment efficacy; and • prepare and implement a clinical follow-up plan. See Figure 3.1 for flow of persons through all nine stages in the management of radionuclide contamination. Warning: Some of the drugs described in this section are quite toxic. Providers should thoroughly familiarize themselves with the latest information on any drug described in this section before prescribing it. Note: Routine or emergency guidance on the use of this section may be obtained by calling REAC/TS at (865) 576-1005. 176

12.1 INTRODUCTION

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12.1 Introduction Internal contamination can occur anytime radioactive materials are free to disperse in an environment. Various documents (Bhattacharyya et al., 1992; Goans, 2002; Gusev et al., 2001; Henge-Napoli et al, 2000; Mettler and Upton, 1995; Ricks et al., 2002) give an overview of current thoughts on the medical management of internal contamination. The early literature on treatment of internal contamination with radionuclides with chelating agents, drugs, and dietary manipulation is accessible in the bibliographies of the comprehensive reviews by Catsch (1964), NCRP (1980), and Volf (1978). A number of activities in the nuclear-fuel cycle (i.e., mining, processing, fabrication of fuel elements, reactor operations and repair, decommissioning, fuel reprocessing, waste management) or in other industrial processes may result in unintentional releases of radioactive material despite engineering and radiation-safety controls. Section 17 describes numerous settings in which radionuclides may be dispersed into the environment with potential human exposures. The most common routes of internal contamination in industrial settings are inhalation and absorption through wounds. Release of radioactive materials may also occur with malicious intentional acts (e.g., terrorist activities in medical institutions, research facilities, public areas, or the environment). Although the ingestion pathway is an uncommon route of contamination in the workplace, it may become critical for the general public after a release of airborne or liquid radioactive material into the environment or deliberate contamination of the food/water supply. Following an intake of radioactive material, the internal dose, toxicity, and treatment methods will depend on factors such as the identity of the radionuclide and its physical and biological characteristics (radioactive half-life, particle size, chemical composition, solubility, and biokinetics). In the inhalation pathway, particle characteristics (size, chemical composition, and solubility in body fluids) are important determinants of dose. The size of aerosol particles and respiratory characteristics such as mouth versus nose breathing and physical activity determine the amount deposited and the regions of the respiratory tract where inhaled particles will be deposited. Clearance from the sites of deposition occurs by two mechanisms (modeled as competing processes), particle transport by physical processes and absorption into blood with transport to other tissues in the body. Physical processes involve mucociliary action moving mucusbound particles and particles taken up by macrophages from various regions of the lungs to thoracic lymph nodes or from the

178 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL) respiratory tract to the esophagus where they may be swallowed and passed through the GI tract, similarly to ingested particles. Highly-insoluble particles may remain in pulmonary tissues for long periods of time; particles transported to the thoracic lymph nodes may eventually reach concentrations that exceed the concentrations in pulmonary tissues. The second clearance mechanism involves dissolution of particulate material and absorption into circulating blood. It is characteristic of even the most insoluble materials that a fraction of the amount inhaled will be immediately dissolved and absorbed into the blood. Early decorporation treatment with chelating agents, for example, can be effective in removing this rapidly absorbed radionuclide. The amount immediately absorbed and subsequent dissolution and absorption processes are a function of the solubility of the inhaled particles. The ICRP respiratory tract model assigns 100, 10 and 0.1 % as the amounts rapidly absorbed for Types F, M and S materials, respectively. Thus, for Type F materials, clearance is entirely by absorption, while for Types M and S, clearance is by both particle transport and absorption (ICRP, 1994a; 2002a) (see Section 16.4.1.2 for additional information). In nearly all cases involving intakes of radionuclides, the absorption from the respiratory tract, GI tract, and wounds, as well as the subsequent biokinetics, will be similar to those of the stable element having the same physical-chemical properties. However, in contrast to stable elements, the elimination of radioactive elements from tissues is affected by radioactive decay as well as biological processes. The concept of an effective half-life was developed for radiation dosimetry modeling to take into consideration these two processes in describing the rate of exponential elimination of radioactive elements from the body. The effective half-life is a function of the biological elimination of a radionuclide from the body, expressed as its biological half-time (Tbiol), and its radioactive decay, expressed as its radioactive half-life (T1/2). These act in parallel to give the effective half-life (Teff) of a radionuclide: (T T ) ( T biol + T 1/2 )

biol 1/2 Τ eff = ---------------------------------

(12.1)

The effective half-life is less than either the radioactive half-life or biological half-time. For example, radioactive cesium acts chemically the same as the nonradioactive isotope of cesium. Therefore, the biological clearance of radiocesium or normal cesium is identical, except for the inclusion of radioactive decay. Although the use of the effective half-life concept has been replaced by rate constants in current biokinetic models for radiation dosimetry purposes

12.2 DECORPORATION THERAPY

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(Sections 16.3 and 20), it is nevertheless useful in describing the retention of radionuclides in the body. Section 16 provides further information on radionuclide radiation biology. General medical and health-physics assessments in an inhalation accident or incident should include initial attempts to determine the maximum credible potential health impact. The assessment of external and internal radionuclide contamination and estimating radiation doses are described in Sections 7, 10 and 21. Section 11 offers guidance to medical personnel in making treatment decisions by introducing an operational unit, the Clinical Decision Guide (CDG). Section 19 describes instrumentation used in assessing radionuclide intakes and Section 20 describes the dosimetric models that are the basis for estimating radiation doses from internal radionuclide contamination. The latter is particularly relevant to this section on mitigating the health effects of radionuclide intakes. Case studies in Section 20 describe treatments and their outcomes for a number of radionuclide contamination incidents. Treatment considerations for internal depositions of radionuclides fall into several major categories (NCRP, 1980): • reduce and/or inhibit absorption of the radionuclide from the GI tract (e.g., induce emesis, gastric lavage, cathartics); • block uptake to the organ of interest (e.g., administer KI to block uptake of radioactive iodine by the thyroid); • utilize isotopic dilution (e.g., increase fluid hydration for internal tritium contamination); • alter the chemistry of the substance (e.g., prevent deposition of uranium carbonate complexes in the renal tubules by use of sodium bicarbonate); • displace the isotope from receptors (e.g., administer iodine to displace 99mTc); • utilize chelation techniques (e.g., administer DTPA for internal deposition of plutonium); • excise radionuclides from wounds; and • consider use of bronchoalveolar lavage for insoluble inhaled particles. 12.2 Decorporation Therapy for Internally-Deposited Radionuclides Since intakes of radionuclides are often considered by physicians as a type of poisoning, poison control centers are occasionally contacted for information on reducing the body burden of the specific radioactive element. It is therefore important that toxicologists

180 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL) and all physicians in poison control centers have access to the most current treatment protocols. Hospital emergency personnel should expect to triage persons exposed and/or contaminated in a radiation incident using traditional medical and trauma criteria. Early identification of the radionuclide is crucial in the medical management of the acute phase. From medical experience with unintentional exposures from contamination in industrial radiation settings, decorporation therapy is generally recommended for intakes equivalent to 1 to 2 CDG (Section 11), and certainly for intakes exceeding 2 CDG. It would be unusual to treat a nonpregnant adult below ~0.2 CDG. However, if the patient requests treatment and if there are sufficient resources available, physicians may wish to treat below this level of intake. For some drugs, FDA specifies not only an indication but also an action level. For those drugs (e.g., DTPA, KI, and Prussian blue), FDA guidance overrides the concept of CDG. In the United States, there are relatively few drugs approved by FDA for the removal of radionuclides from the body. This problem exists worldwide. The drugs that are approved as chelating agents or decorporation agents in the United States are listed in Table 12.1. Uses of these drugs for other indications must be considered “off-label.” This section provides guidance in two ways: (1) recommendations for decorporation therapy for specific radionuclides and (2) drug information for treatment. Table 12.2 presents a summary of treatment recommendations for various radionuclides of concern in the medical and industrial environment. Table 12.3 provides dose schedules for drug or treatment modalities. Decorporation therapy by drugs and detailed treatment for selected radionuclides are described in Sections 12.3 and 12.4. Lung lavage is described in Section 12.5. Treatment of contaminated wounds is not specifically addressed in this Report with the exception of plutonium and other actinides in Section 12.3.3. Readers are referred to NCRP Report No. 156 (NCRP, 2006a) for extended information on the behavior in wounds of various radionuclides in different physical and chemical forms and the associated dosimetric and treatment information. The majority of the drugs listed in Table 12.2 and discussed throughout this section, are not approved by FDA for the indications listed. Major references for these uses include Bhattacharyya et al. (1992), Henge-Napoli et al. (2000), NCRP (1980), and Stradling and Taylor (2005), and clinical experience at REAC/TS.8 There is limited 8Wiley,

A., Jr. (2008). Personal communication (Radiation Emergency Assistance Center/Training Site, Oak Ridge, Tennessee).

12.2 DECORPORATION THERAPY

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TABLE 12.1—Drugs approved for chelation or decorporation by FDA.a Drug

Acetylcysteine [N-acetyl-L-cysteine (NAC)] [Acetadote® (Cumberland Pharmaceuticals, Nashville, Tennessee)], injection Deferasirox [Exjade® (Novartis Pharma, Stein, Switzerland)], oral Deferoxamine (DFOA) [Desferal® (Novartis Pharma, Stein, Switzerland)], injection Dexrazoxane [Zinecard® (Pfizer, Inc., New York)], injection Dimercaprol (British Anti-Lewisite), injection Diethylenetriaminepentaacidic acid (DTPA), injection Ethylenediaminetetraacetic acid (EDTA) (Edetate Calcium Disodium) [Calcium Disodium Versenate® (Graceway Pharmaceuticals, Bristol, Tennessee)], injection Penicillamine [Cuprimine® (Merck and Company, Inc., Whitehouse Station, New Jersey)], oral

Potassium iodide (KI), oral Prussian blue [insoluble ferric hexacyanoferrate (II)] [Radiogardase® (Heyltex Corporation, Katy, Texas)], oral Sevelamer HCL [Renagel® (Genzyme Corporation, Cambridge, Massachusetts)], oral Succimer [dimercaptosuccinic acid (DMSA)] (Chemet®), oral

FDA Indication

To prevent or lessen hepatic injury after ingestion of a potentially-hepatoxic quantity of acetaminophen (Iron) Chronic iron overload (Iron) Acute iron intoxication and chronic iron overload Cardiomyopathy associated with doxorubicin administration in women with metastatic breast cancer Arsenic, gold, and mercury poisoning and acute lead poisoning when used concomitantly with EDTA Plutonium, americium, curium (Lead) Reduce blood lead levels and depot stores in lead poisoning (acute and chronic) and lead encephalopathy

(Copper) Treatment of Wilson’s disease (excess copper accumulation), cystinuria, and in patients with severe, active rheumatoid arthritis who have failed to respond to conventional therapy Radioactive iodine Radioactive cesium and/or radioactive or nonradioactive thallium

(Phosphorus) Control of serum phosphorus in patients with chronic kidney disease on hemodialysis (Lead) Treatment of lead poisoning in pediatric patients with blood lead levels >45 μg dL–1 ® Trientine HCL [Syprine (Merck and (Copper) Treatment of Wilson’s Company, Inc., Whitehouse Station, disease (excess copper accumulation) New Jersey)], oral in patients who are intolerant of penicillamine a All drugs listed in this table, except KI, are prescription drugs (FDA, 2009). KI is available over-the-counter. DTPA, KI and Prussian blue are available from the Strategic National Stockpile (CDC, 2008).

Radionuclides

Possible Treatments

Preferred Prescription

Actinium

Consider DTPA

Consider DTPA

Americium

DTPA

DTPA

Antimony

British Anti-Lewisite (BAL), penicillamine

BAL

Arsenic

BAL, dimercaptosuccinic acid (DMSA)

BAL

Barium

Barium, calcium therapy (Section 12.4.1)

Section 12.4.1

Berkelium

DTPA

DTPA

Bismuth

BAL, Penicillamine, DMSA

DMSA

Cadmium

DMSA, DTPA, Ethylenediaminetetraacetic acid (EDTA)

DMSA

Californium

DTPA

DTPA

Calcium

Barium, calcium therapy (Section 12.4.1)

Section 12.4.1

Carbon

Consider hydration and nonlabeled carbon

Consider hydration and nonlabeled carbon

Cerium

DTPA

DTPA

Cesium

Prussian blue

Prussian blue

Chromium

DTPA, EDTA (antacids are contraindicated)

DTPA

Cobalt

DMSA, DTPA, EDTA, N-acetyl-L-cysteine (NAC)

DTPA

Copper

EDTA, penicillamine, trientine

Penicillamine

Curium

DTPA

DTPA

Einsteinium

DTPA

DTPA

Europium

DTPA

DTPA

Fission products (mixed)

Management depends on predominant radionuclides present at the time (e.g., early: iodine; late: strontium, cesium, and others)

182 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL)

TABLE 12.2—Decorporation therapy recommendations in the United States for radionuclides of concern.a

Aluminum hydroxide

Aluminum hydroxide

Gallium

Consider penicillamine

Penicillamine

Gold

BAL, penicillamine

BAL

Indium

DTPA

DTPA

Iodine

KI, consider SSKI, propylthiouracil, methimazole or potassiumiodate

KI

Iridium

Consider DTPA, EDTA

Consider DTPA

Iron

Deferoxamine (DFOA), deferasirox, DTPA, DFOA and DTPA together

DFOA

Lanthanum

DTPA

DTPA

Lead

DMSA, EDTA, EDTA with BAL

DMSA

Manganese

DFOA, DTPA, EDTA

DTPA

Magnesium

Consider strontium therapy (Section 12.4.5)

Consider strontium therapy

Mercury

BAL; EDTA; penicillamine; DMSA

BAL

Molybdenum

Limited clinical experience

Neptunium

Consider DFOA and/or DTPA

Consider DFOA and/or DTPA BAL

BAL, EDTA DTPA

DTPA

Palladium

Penicillamine, DTPA

Penicillamine

Phosphorus

Phosphorus therapy (Section 12.4.4)

Phosphorus therapy

Plutonium

DTPA, DFOA, EDTA, DTPA and DFOA together

DTPA

Polonium

BAL, DMSA, penicillamine

BAL

Potassium

Diuretics

Diuretics

Promethium

DTPA

DTPA

Radium

Radium, strontium therapy (Section 12.4.5)

Section 12.4.5

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Nickel Niobium

12.2 DECORPORATION THERAPY

Fluorine

Radionuclides

Possible Treatments

Preferred Prescription

Rubidium

Prussian blue

Prussian blue

Ruthenium

DTPA, EDTA

DTPA

Scandium

DTPA

DTPA

Silver

No specific therapy

Sodium

Diuretic and isotopic dilution with 0.9 % NaCl

Diuretic and isotopic dilution with 0.9 % NaCl

Strontium

Radium, strontium therapy (Section 12.4.5)

Section 12.4.5

Sulfur

Consider sodium thiosulfate

Consider thiosulfate

Technetium

Potassium perchlorate

Potassium perchlorate Prussian blue

Thallium

Prussian blue

Thorium

Consider DTPA

Consider DTPA

Tritium (3H)

Force fluids

Water diuresis

Uranium

Bicarbonate to alkalinize the urine; consider dialysis

Bicarbonate

Yttrium

DTPA, EDTA

DTPA

Zinc

DTPA, EDTA, zinc sulfate as a diluting agent

DTPA

Zirconium

DTPA, EDTA

DTPA

aThe

majority of these drugs are not approved by FDA for the indications listed in this table. Major references for this table include Bhattacharyya et al. (1992), Henge-Napoli et al. (2000), and NCRP (1980), and clinical experience at REAC/TS. There is limited clinical experience with many of these recommendations. See individual drug labels for detailed guidance on these drugs.

184 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL)

TABLE 12.2—(continued)

TABLE 12.3—Dose schedules for drug or treatment modalities.a,b Drug or Treatment Modality

Dosage

Deferoxamine (DFOA) mesylate (Section 12.3.1)

FDA does not specify age: DFOA mesylate injectable; IM is preferred. 1 g IM or IV (2 ampules) slowly (15 mg kg –1 h–1); Repeat as indicated as 500 mg IM or IV q4h × 2 doses; then 500 mg IM or IV every 12 h for 3 d.

Dimercaprol (British Anti-Lewisite) (Section 12.3.2)

FDA does not specify age: IM: 300 mg per vial for deep IM use, 2.5 mg kg –1 (or less) every 4 h for 2 d, then twice daily for 1 d then daily for days 5 – 10.

Diethylenetriaminepentaacidic acid (DTPA, calcium or zinc) (pentetate calcium trisodium and Pentetate zinc trisodium) (Section 12.3.3)

Adults: IV: 1 g in 5 mL IV push over 3 to 4 min or IV infusion over 30 min diluted in 250 mL of 5 % dextrose in water, Ringers lactate or normal saline. Nebulized inhalation: 1g in 1:1 dilution with sterile water or normal saline. Children under 12 y: 14 mg kg –1 IV as above, not to exceed 1 g.

Edetate calcium disodium [Ethylenediaminetetraacetic acid (EDTA)] Section 12.3.4)

FDA does not specify age: Ca-EDTA (edetate calcium disodium); 1,000 mg m–2 d–1 added to 500 mL 5 % dextrose or 0.9 % sodium chloride infused over 8 to 12 h. This same dosage can be given IM divided into equal doses spaces 8 to 12 h apart.

Penicillamine (Section 12.3.5)

FDA does not specify age: Oral: 250 mg daily between meals and at bedtime. May increase to 4 or 5 g daily in divided doses.

Phosphorus therapy Potassium phosphate, dibasic (Section 12.4.4)

Oral: 250 mg phosphorus per tablet. Adults: 1 – 2 tablets oral four times daily with full glass of water each time, with meals and at bedtime. Children >4 y of age: 1 tab oral four times daily.

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Consider dosage as for acetaminophen overdosage, start at 140 mg kg –1 oral loading dose (RxList, 2009).

12.2 DECORPORATION THERAPY

Acetylcysteine [N-acetyl-L-cysteine (NAC)] (Section 12.4.2)

Drug or Treatment Modality

Dosage

Potassium iodide (KI) (Section 12.4.3)

Oral: tablets or liquid. Drug dose varies between 16 and 130 mg daily depending on age, thyroid exposure level, and whether or not pregnant or lactating (Table 12.14).

Propylthiouracil (Section 12.4.3)

Oral: 50 mg tablets, 2 tablets three times daily for 8 d. FDA does not specify age.

Prussian blue (Section 12.3.6)

Oral: Adults and adolescents 3 g three times daily. Children 2 to 12 y of age: 1 g three times daily.

Sodium bicarbonate (for uranium only) (Section 12.4.7)

Oral or IV (Table 12.22).

Radium and strontium therapy (Section 12.4.5)

Section 12.4.5.

Succimer [dimercaptosuccinic acid (DMSA)] [Chemet® (Schwarz Pharma, Monheim, Germany)] (Section 12.3.7)

FDA approved pediatric dosing: Start dosage at 10 mg kg –1 or 350 mg m–2 oral every 8 h for 5 d. Reduce frequency of administration to 10 mg kg –1 or 350 mg m–2 every 12 h (two-thirds of initial daily dosage) for an additional two weeks of therapy. A course of treatment lasts 19 d.

Water diuresis (Section 12.4.6)

Oral: Fluids >3 – 4 L d–1.

aUnless b

noted otherwise, the references for these dose schedules are given in the listed sections. Dosage notations: IV = intravenous injection bid = twice per day IM = intramuscular injection tid = three times per day mEq = milliequivalent qid = four times per day PO = per os or orally qd = every day q12h = every 12 h q4h = every 4 h

186 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL)

TABLE 12.3—(continued)

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clinical experience with many of these recommendations. See individual drug labels for detailed guidance on these drugs. Every drug approved by FDA for usage in the United States is assigned a pregnancy category. These categories help characterize risk of drug effect on the unborn infant. This risk should be considered in the total benefit-to-risk picture of possible injury to a patient and the unborn from an internal radionuclide versus benefit from a drug or other treatment which may be toxic to both the patient and the unborn. Pregnancy categories are defined in Appendix J. 12.3 Decorporation Therapy by Drug 12.3.1 Deferoxamine Treatment [Deferoxamine (DFOA) mesylate, also called desferrioxamine, Desferal®]. FDA Indication: Treatment of acute iron intoxication and chronic iron overload due to transfusion dependent anemia. Radionuclide Decorporation Therapy: FDA indication: iron. Not approved by FDA but considered by other authors: manganese, neptunium, plutonium. Notes: 1. The use of DTPA for neptunium is controversial. Volf and Wirth (1986) described reduced retention of 239Np in rats by early combined treatment with DTPA and DFOA. Stradling and Taylor (2005) reported that DTPA was ineffective with neptunium. 2. NCRP Report No. 65 (NCRP, 1980) states that “DFOA equals or surpasses CaDTPA in enhancing the excretion of plutonium (IV) compounds provided the drug is given promptly…. The combination of DFOA and CaDTPA yields better results than either drug separately in the treatment of plutonium poisoning.” Modes of Treatment: Deferoxamine (DFOA) mesylate injection. IM preferred: 1 g. 1 g IV (two ampules) slowly (15 mg kg –1 h–1). Administer once; then obtain urine and fecal bioassays to assess. Repeat as indicated as 500 mg IM (preferred) or IV q4 h × 2 doses; then 500 mg IV q12 h for 3 d (Table 12.4).

Administer

Notes

Deferoxamine (DFOA) mesylate injectable; IM is preferred. 1 g IM or IV (2 ampules) slowly (15 mg kg –1 h–1)

DFOA binds iron in ferritin. Promotes excretion of iron and inhibits absorption

Repeat as indicated as 500 mg IM or IV q4h × 2 doses; then 500 mg IM or IV q12h for 3 d

Pregnancy Category Cb

Titrate on mg kg –1 basis for children if on extended therapy a

FDA (2009). categories are defined in Appendix J.

bPregnancy

Contraindications

Severe renal disease or anuria

Monitor Clinically

Rapid infusion may lead to hypotension and shock

Administer once; then obtain urine and fecal bioassay to assess efficacy of treatment

188 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL)

TABLE 12.4—DFOA treatment.a

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Contraindications: Severe renal disease or anuria. Rapid infusion may lead to hypotension and shock. Pregnancy: Category C (Meadows, 2001). 12.3.2 Dimercaprol Treatment [British Anti-Lewisite (BAL)] toxic, seldom first drug of choice. Succimer [dimercaptosuccinic acid (DMSA)] is less toxic and may be of equal efficacy. Radionuclide Decorporation Therapy: FDA indication: Treatment of arsenic, gold and mercury poisoning. It is indicated in acute lead poisoning when used concomitantly with edetate calcium disodium injection (U.S. Pharmacopeia, Rockville, Maryland) (FDA, 2009). Not approved by FDA but considered by other authors: antimony, bismuth, chromium, nickel, polonium. Notes: 1. NCRP Report No. 65 (NCRP, 1980) states that dimercaprol may be considered for treatment of internal contamination with arsenic, bismuth, chromium, gold, lead, mercury and nickel and it also says that dimercaprol can be considered for polonium. 2. Bhattacharyya et al. (1992) lists BAL for treating internally-deposited arsenic, gold, bismuth, nickel, lead, polonium, and antimony. 3. Flomenbaum et al. (2006) states that BAL may be beneficial for bismuth poisoning. Modes of Treatment: BAL – Deep IM injection only, 2.5 mg kg –1 qid for 2 d, bid third day, then daily for 5 to 10 d (Table 12.5). Contraindications: Hepatic insufficiency or acute renal insufficiency; peanut allergy (peanut oil suspension). Precautions: BAL is a highly toxic drug, and the unknown benefitto-risk ratio for off-label uses warrants extreme caution. Alkaline urine protects the kidneys during therapy. Hypertension and tachycardia are common. There may be a need for antihistamine therapy. BAL may cause a sterile abscess. Pregnancy: Category C (Meadows, 2001).

Administer –1

Notes

BAL – 2.5 mg kg IM qid for 2 d, bid third day, then daily for 5 – 10 d

BAL forms metal chelates with many metals

How supplied: 3 mL (100 mg mL–1) ampules

First test for hypersensitivity with 0.25 ampule Pregnancy Category Cb

aFDA b

(2009). Pregnancy categories are defined in Appendix J.

Contraindications

Hepatic insufficiency; acute renal insufficiency.

Monitor Clinically

Clinically-significant adverse reactions may follow large, though not extra-therapeutic doses These include: fatigue, weakness, paresthesia, lacrimination, vomiting, tachycardia and hypertension

190 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL)

TABLE 12.5—Dimercaprol [British Anti-Lewisite (BAL)] treatment (for arsenic, bismuth, chromium, gold, lead, mercury, nickel, and polonium).a

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12.3.3 DTPA Treatment [Pentetate calcium trisodium injection and pentetate zinc-trisodium injection, also known as trisodium calcium and trisodium zinc diethylentriaminepentaacetate (Ca-DTPA and Zn-DTPA) (FDA, 2009)]. FDA indication: Treatment of individuals with known or suspected internal contamination with plutonium, americium or curium to increase the rates of elimination. Radionuclide Decorporation Therapy: FDA indication: americium, curium, plutonium. Not approved by FDA but considered by other authors: actinium, berkelium, cadmium, californium, cerium, chromium, cobalt, einsteinium, europium, indium, lanthanum, manganese, neptunium, niobium, promethium, ruthenium, scandium, thorium, yttrium, zinc, zirconium. Notes: 1. There is controversy in the above list of radionuclides. Stradling and Taylor (2005) state that DTPA is not effective for treating internal contamination due to neptunium, thorium or uranium but is effective for treating other actinides. 2. NCRP Report No. 65 (NCRP, 1980) states that DTPA is effective for neptunium, and also californium, cerium, chromium, europium, indium, lanthanum, niobium, promethium, scandium, yttrium, zinc and zirconium. 3. The report by Henge-Napoli et al. (2000) states in different sections that DTPA is the recommended chelate for thorium, but also that the effectiveness of DTPA with thorium is poor. 4. Limited literature describes use of DTPA off-label for treatment of cobalt (Section 12.4.1) and cadmium, chromium, manganese and ruthenium (Bhattacharyya et al., 1992). 5. Flomenbaum et al. (2006) note that DTPA has been shown to be effective for cadmium and zinc poisoning in animal studies. 6. Leikin and Paloucek (2008) note that DTPA may be useful for removing californium, cerium, lanthanum, lutetium, manganese, niobium, promethium, scandium, thorium, uranium, yttrium, zinc and zirconium.

192 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL) Summary of Treatment Recommendations Modes of Treatment [intravenous or inhalation administration (FDA, 2009)]: Adults and adolescents: Start: Ca-DTPA, 1 g in 5 mL slow IV push over 3 to 4 min first day (Table 12.6). Maintenance: Zn-DTPA, 1 g in 5 mL slow IV push daily over 3 to 4 min. Or DTPA IV in 100 to 250 mL 5 % dextrose in water, Ringers Lactate, or normal saline infused over 30 min. Use Ca-DTPA the first day then change to Zn-DTPA. Or by nebulized inhalation if the only contamination is through inhalation, 1 g DTPA in 5 mL diluted 1:1 with sterile water or saline. Treatment of Contaminated Wounds: Irrigation of wounds with DTPA may be considered for some radionuclides such as the lanthanides and actinides (NCRP, 2006a). Ca- or Zn-DTPA can be used in these cases. Early irrigation with DTPA may enhance the effectiveness of the irrigation process and by binding the metals into stable complexes reduce tissue uptake and allow urinary excretion. An effective irrigating solution consists of 1 g Ca- or Zn-DTPA and 10 mL of 2 % lidocaine in 100 mL of 5 % glucose solution or sterile isotonic saline (NCRP, 1980). The irrigation can also be accompanied by IV and inhalation administration of DTPA, but care must be taken to avoid overdosing with DTPA because the amount absorbed by the wound tissues cannot be measured. Further discussion of decontaminating wounds appears in Sections 8.1.2 and 9.4.1. Pediatrics: Start Ca-DTPA, then change to Zn-DTPA as for adults. Dose: 14 mg kg –1 IV, not to exceed 1 g d–1 (children <12 y of age). Nebulized inhalation is not approved for pediatric use. Administer one dose in mass casualty incidents; the first dose is the most important for chelation. Obtain urine bioassay to assess efficacy of the first dose and as a basis to consider further therapy. Contraindications: None. However see “Use in Pregnancy” below. Efficacy: Monitor blood pressure during administration; Ca-DTPA is ~10 times more effective than Zn-DTPA within the first 24 h; DTPA can reduce dose by 80 % for soluble forms if given within 24 h, but <25 % after intake of soluble or insoluble compounds. The first dose is the most important; most persons receiving DTPA for unintentional internal contamination have received only one dose.

TABLE 12.6—Ca- or Zn-DTPA treatment.a Contraindications

Monitor Clinically

First day administer Ca-DTPA, then for maintenance use Zn-DTPA. FDA approved alternate modes of administration:

Mechanism of action; heavy metal chelation.

Ca-DTPA is contraindicated in the nephritic syndrome or in cases of renal insufficiency or renal failure.

Blood pressure during administration.

Ca- or Zn-DTPA, 1 g slow IV push over 3 – 4 min. Also, Ca- or Zn-DTPA 1 g IV in 100 – 250 mL 5 % dextrose in water, Ringers lactate or normal saline infused over 30 min.

Administer 1 dose in mass casualty incidents; the first dose is the most important for chelation efficacy. Commonly, most contaminated patients have received only one dose IV.

Ca-DTPA has Pregnancy Category C,b Zn-DTPA has Pregnancy Category B. Therefore, only Zn-DTPA should be utilized in pregnancy.

Occasional local skin reaction may occur.

Also, adults only may use nebulized inhalation, 1 g DTPA diluted at a 1:1 ratio with sterile water or saline.

Obtain bioassay to assess efficacy of the first dose and as a basis to consider further therapy.

DTPA is not recommended for chelation with uranium.c

Rare dizziness during administration has been noted but is transient due to blood pressure fluctuations.

For children, <12 y of age, 14 mg kg –1 as above

Ca-DTPA is ~10 times more effective than Zn-DTPA within the first 24 h.

Additional rare side effects: nausea, diarrhea, metallic taste, headache, and chest pain.

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Notes

12.3 DECORPORATION THERAPY BY DRUG

Administer

Administer

Notes

DTPA can reduce dose by 80 % if given within 24 h, but <25 % after intake of insoluble compounds.

aFDA

(2009). Pregnancy categories are defined in Appendix J. cHenge-Napoli et al. (2000). b

Contraindications

Monitor Clinically

Minor trace elements such as manganese, magnesium and zinc may be chelated with Ca-DTPA. These should be monitored or the patient given vitamin and mineral supplements as indicated.

194 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL)

TABLE 12.6—(continued)

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Use in Pregnancy and Nursing Mothers: Ca-DTPA is Pregnancy Category C, Zn-DTPA is Pregnancy Category B (Meadows, 2001). Treatment of pregnant women should begin and continue with Zn-DTPA. Because animal reproduction studies are not always predictive of human response, this drug should be used during pregnancy only if clearly needed. The risk of toxicity from untreated internal radioactive depositions should be weighed against the risk of Zn-DTPA treatment. Studies to determine if Zn-DTPA is excreted in breast milk have not been conducted, but CaNa3-DTPA distribution in extracellular water can be expected to appear in milk. Many radionuclides are known to be excreted in varying degrees in breast milk. Women with known or suspected internal depositions of radionuclides should not breast feed, whether or not they are receiving chelation therapy. Renally-Impaired Patients: No dose adjustment is needed. However, renal impairment may reduce the rate at which chelation removes radionuclides from the body. In heavily contaminated patients with renal impairment, dialysis may be used to increase the rate of elimination. High-efficiency, high-flux dialysis is recommended. Because dialysis fluid will become radioactive, radiation precautions must be taken to protect personnel, other patients, and the general public. Duration of Maintenance Treatment: The duration of chelation treatment depends on the amount of internal deposition and individual response to treatment. Monitoring: When possible, obtain baseline blood and urine samples (CBC with differential, bound urinary nitrogen, serum chemistries, and electrolytes, urinalysis and blood and urine bioassay) before initiating treatment. Ca-DTPA must be given with very careful monitoring of serum zinc and complete blood counts. To establish an elimination curve, a quantitative baseline estimate of the total internalized transuranium element(s) and measures of elimination of activity should be obtained by appropriate wholebody counting or by appropriate serial bioassay. General Information on DTPA Chelation Therapy for Actinide Radionuclides: The actinides are 15 heavy elements at the bottom of the Periodic Table of the Elements beginning with actinium [atomic number (Z) = 90] and ending with lawrencium (Z = 103).

196 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL) New transactinides have different chemical properties and their half-lives are so short they are not a contamination problem. The primary actinides for consideration in internal radionuclide depositions are plutonium, americium, curium and californium. Plutonium-238, 239Pu and 240Pu are the most commonly encountered plutonium isotopes, while 241Am and 244Cm are the most commonly encountered isotopes of americium and curium. All have long radioactive and biological half-times, and inhalation exposures account for ~75 % of industrial contamination incidents. Plutonium may be considered as a qualitative model for the biological behavior of the actinide elements [with the exceptions of uranium and neptunium (V), which have different patterns of distribution and excretion]. If an inhaled plutonium compound is relatively soluble (e.g., nitrate or citrate), a substantial portion of the activity deposited in the respiratory tract is absorbed directly to blood. Roughly 80 % of plutonium reaching blood deposits in the liver and bone over a period of hours or days. The division between liver and bone is variable, but current evidence suggests that, on average, the liver contains more plutonium than the skeleton in adults within the first year after contamination. Over a period of years, bone becomes the dominant repository for systemic plutonium due to more tenacious retention of plutonium in bone than in liver. The rest of the activity is distributed throughout other tissues except for a small percentage that is filtered by the kidneys or secreted into the GI contents and passed in the feces. Plutonium is released from bone, liver, and most other tissues back to blood with half-times of at least a few years. It is then apparently recycled to tissues and excretion pathways in roughly the same ratios as the initial input to blood. This pattern of recycling results in a net removal half-time from the human body of several decades. Urine bioassay using alpha spectrometry or thermal ionization mass spectrometry is currently the measurement of choice for contamination with plutonium. Fecal bioassay samples are useful in accidental and emergency situations, and are typically analyzed with alpha spectrometry. In vivo measurement of plutonium is difficult for common mixtures of plutonium isotopes due to the low yield and energy of photon emissions. The level of internally deposited plutonium often is estimated on the basis of external measurement of photons from 241Am that generally accompanies plutonium in the work place or is produced in the body by decay of 241Pu (Section 20.15). In vivo measurements of 241Am are accomplished using HPGe thin detectors in shielded rooms (Table 19.1). External measurements of 239Pu or 241Am is feasible for wound monitoring

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because typically there is little attenuation by overlying soft tissue for most wounds (Section 19.2.7.1). Ca- and Zn-DTPA may be considered for chelation therapy for internal contamination of all the actinides except thorium, uranium and neptunium (Stradling and Taylor, 2005). Ca-DTPA is a calcium complex of DTPA and Zn-DTPA is similar, except for the substitution of zinc for calcium. These compounds have been widely used in the United States and in Europe as chelating agents for plutonium and other transuranic elements such as americium, californium and curium. DTPA is a synthetic poly-amino polycarboxylic acid that forms stable complexes (metal chelates) with a large number of metal ions. The drug effectively exchanges calcium or zinc for another metal of greater binding power and carries it to the kidneys where it is excreted into the urine. The plasma half-life of DTPA is 20 to 60 min, and almost the entire administered dose is excreted in 12 h. However, clinical experience as documented in the REAC/TS Registry has shown that in some cases modest enhancement of urinary excretion of plutonium has been observed months to years after an incident. Ca-DTPA is ~10 times more effective than Zn-DTPA for initial chelation of transuranics within 24 h of contamination. Therefore, Ca-DTPA should be used whenever large body burdens of transuranics are involved. The chelating efficacy is greatest immediately or within 1 h of intake. Approximately 24 h after the contamination incident, Zn-DTPA is, for all practical purposes, as effective as CaDTPA. This comparable efficacy, coupled with its lesser toxicity, makes Zn-DTPA the preferred agent for protracted therapy. If Ca-DTPA is not available, treatment should start with Zn-DTPA. Clinical Experience: REAC/TS at Oak Ridge Associated Universities maintains a DTPA Registry for DOE. DTPA use is monitored for FDA, and DTPA is maintained by a number of co-investigators throughout the United States and the world. Since 2004, DTPA has been approved by FDA for treatment of individuals with known or suspected internal depositions of plutonium, americium or curium to increase the rates of elimination. DTPA is stocked by the U.S. Strategic National Stockpile of CDC. Additional information on DTPA is available on the FDA website (FDA, 2009). Over the past 50 y, an adverse reaction rate of ~2.7 % has been noted with DTPA. Most reactions have been minor and the drug is considered quite safe in a highly selected patient population of healthy workers. Adequate and well-controlled pharmacokinetic and pharmacodynamic studies in renally-impaired and/or hepaticallyimpaired patients have not been identified in the literature. Both

198 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL) DTPA and its radioactive chelates are excreted by glomerular filtration. Impaired renal function may decrease rates of elimination and increase the serum half-life of DTPA. In the DTPA Registry, a total of 646 individuals received at least one dose of either Ca- or Zn-DTPA. Of these, 62 received Zn-DTPA by one or more routes of administration. Forty-eight individuals were dosed by intravenous administration, 18 by inhalation, and eight by other or unknown routes of administration. Of the individuals that received Zn-DTPA, 23/62 (37 %) received one dose and eight (13 %) received two doses. The remaining 31 individuals received three or more doses (FDA, 2009). The largest number of Zn-DTPA doses given to a single individual was 574 doses delivered over 3.5 y. Overall, the presence or absence of adverse incidents was recorded in 310/646 individuals. Of these, 19 (6.1 %) individuals reported at least one adverse incident. The total number of recorded adverse incidents was 20. Of the 20 adverse incidents, one individual treated with Zn-DTPA reported headache, light headiness, and pelvic pain. Two individuals experienced cough and/or wheezing with nebulized Ca-DTPA therapy. However, there was no report of such incidents with nebulized Zn-DTPA (FDA, 2009). Nebulized chelation therapy may be associated with exacerbation of asthma. Therefore, caution should be exercised when administering DTPA by the inhalation route. Treatment over several months with DTPA can lead to depletion of endogenous metals (e.g., magnesium, manganese, zinc). These elements should be monitored routinely and, if appropriate, mineral or vitamin plus mineral supplements should be provided. Reports of adequate and well-controlled drug-drug interaction studies in humans are not available in the medical literature. Studies with DTPA to evaluate mutagenesis and impairment of fertility have not been performed. Likewise, data for DTPA effects on spermatogenesis are also not available. 12.3.4 Ethylenediaminetetraacetic Acid Treatment [Ethylenediaminetetraacetic acid (EDTA), Edetate Calcium Disodium Injection, Calcium Disodium Versenate®; do not confuse with edetate disodium, Endrate® (FDA, 2008a)]. FDA Indication: Lead poisoning, lead encephalopathy in pediatric populations and adults. Radionuclide Decorporation Therapy: FDA approved: lead.

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Not approved by FDA but considered by other authors: cadmium, chromium, cobalt, copper, manganese, nickel, zinc. Note: NCRP Report No. 65 (NCRP, 1980) describes the use of EDTA with all the above radionuclides except cobalt. DTPA and EDTA have been shown in laboratory animals to effectively chelate cobalt (Henge-Napoli et al., 2000). Modes of Treatment: Ca-EDTA prescription injection, 1,000 mg m–2 d–1 added to 500 mL 5 % dextrose or 0.9 % sodium chloride infused over 8 to 12 h. This same dosage can be given IM divided into equal doses spaced 8 to 12 h apart. Administer for 1 d as above unless medically indicated for extended treatment (Table 12.7). Contraindications: EDTA should not be given during periods of anuria, nor to patients with active renal disease or hepatitis. Treatment-induced nephrotoxicity has been noted. Urine output should be monitored as well as electrocardiogram changes. Pregnancy: Category B (Meadows, 2001). 12.3.5 Penicillamine Treatment (Cuprimine®) FDA Indication: Treatment of Wilson’s disease, cystinuria, and in patients with severe active rheumatoid arthritis who failed to respond to an adequate trial of conventional therapy. [Wilson’s disease (also known as hepatolenticular degeneration) is a hereditary disease due to excess copper accumulation.] Radionuclide Decorporation Therapy: FDA approved: copper. Not approved by FDA but considered by other authors: bismuth, gold, lead, mercury, polonium. Notes: 1. Leiken and Paloucek (2008) recommend penicillamine or unithiol (not available in the United States) for treatment of bismuth poisoning.

Administer

Notes

1,000 mg m–2 d–1 added to 500 mL 5 % dextrose or 0.9 % sodium chloride infused over 8 to 12 h. This same dosage can be given IM divided into equal doses spaced 8 to 12 h apart. Administer for 1 d as above unless medically indicated for extended treatment.

Pregnancy Category B.b

Contraindications

Monitor Clinically

Ca-EDTA

Edetate calcium disodium is available as 200 mg mL–1 injection. a

RxList (2009). categories are defined in Appendix J.

bPregnancy

Patients with active renal disease or hepatitis.

Urine output and electrocardiogram changes. Administer once; then obtain bioassay to assess efficacy of treatment.

200 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL)

TABLE 12.7—Ca-EDTA treatment.a

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2.

3.

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NCRP Report No. 65 (NCRP, 1980) recommended considering penicillamine for cobalt, gold, lead, mercury and polonium. Bhattacharyya et al. (1992) recommended considering penicillamine for antimony, arsenic, bismuth, copper, gold, and lead internal contamination.

Modes of Treatment: D-penicillamine; 250 mg capsules 0.75 to 1.5 g PO daily for adults and 30 mg kg–1 d–1 divided into four doses for pediatric patients (Table 12.8). Obtain urine and fecal bioassay; continue only if clinically indicated by the magnitude and length of the incident. Contraindications: Nursing mothers, patients with a history of penicillamine-related aplastic anemia or agranulocytosis, and patients with renal insufficiency. Pregnancy: Category D (positive evidence of fetal risk) (Meadows, 2001). 12.3.6 Prussian Blue Insoluble Treatment [Prussian blue insoluble, ferric hexacyanoferrate (II); Radiogardase®]. FDA Indication: Treatment of patients with known or suspected internal contamination of radioactive cesium and/or radioactive or nonradioactive thallium to increase their rates of elimination. Radionuclide Decorporation Therapy: FDA approved: cesium, thallium. Not approved by FDA but considered by other authors: rubidium. Notes: 1. Stather (1972) described the influence of Prussian blue on metabolism of rubidium in rats. 2. Bhattacharrya et al. (1992) recommended use of Prussian blue for rubidium internal contamination. Modes of Treatment: FDA guidelines (FDA, 2009): Adults and adolescents: 3 g orally three times a day. Pediatrics (2 to 12 y of age): 1 g orally three times a day (Table 12.9).

Administer

D-penicillamine; Adults: 0.75 – 1. 5 g oral per day; Pediatrics: 30 mg kg –1 d–1 divided into four doses

Notes

Can react with cystine to form a stable, mixed disulfide. Obtain urine and fecal bioassay to assess efficacy; continue only if clinically indicated by the magnitude and length of the incident. Pregnancy Category D.b

a

FDA (2009). categories are defined in Appendix J.

bPregnancy

Contraindications

Pregnancy unless with Wilson’s disease or cystinuria. Nursing mothers, patients with a history of penicillaminerelated aplastic anemia or agranulocytosis, and patients with renal insufficiency.

Monitor Clinically

High incidence of untoward reactions include rash, fever, lymphadenopathy, thrombocytopenia, nephrotic syndrome, and possibly cataracts and optic neuropathy. May require daily supplement of pyridoxine.

202 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL)

TABLE 12.8—Penicillamine treatment group (for radioactive arsenic, cobalt, copper, gold, lead, mercury, polonium).a

TABLE 12.9—Prussian blue treatment.a Administer

Notes Mechanism of action; ion exchange.

Pediatrics, 2 – 12 y of age: 1 g oral three times daily.

Prussian blue inhibits the enterohepatic cycle in the GI tract, thereby reducing GI half-life. Prussian blue reduces cesium dose by a factor of two to three by inhibiting enterohepatic GI tract recirculation.

Infants (not FDA approved): 0.2 – 0.3 mg kg –1 orally according to limited clinical experience.

Patients should be warned of blue stool and if opened and mixed with food or drink, can stain teeth and gums blue.

Continue for a minimum of 30 d per FDA guidance.

Pregnancy Category C.b

(2009). categories are defined in Appendix J.

Experience with Prussian blue has been entirely in ingestion incidents. Likely not as effective in dose reduction in an inhalation incident. Monitor urine and fecal bioassay to assess treatment efficacy.

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aFDA

bPregnancy

Essentially none in serious incidents. Even though Prussian blue has a cyanide moiety, it is essentially not absorbed from the GI tract. Use with caution in patients with history of GI obstruction and peptic ulcer disease.

Monitor Clinically

12.3 DECORPORATION THERAPY BY DRUG

Adults and adolescents: 3 g oral three times daily.

Contraindications

204 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL) There is no FDA approval for children <2 y of age but consider scaled dose according to limited clinical experience. Monitor urine and fecal bioassay to assess treatment efficacy. Contraindications: None. Pregnancy: Category C (Meadows, 2001). Precautions: Patients should be advised that stool may become blue. Prussian blue may be less effective in patients with impaired liver function due to decreased excretion of cesium and thallium in the bile. Serum electrolytes should therefore be closely monitored during insoluble Prussian blue treatment. Caution should be exercised when treating patients with preexisting cardiac arrhythmias or electrolyte imbalances. Prussian blue may be constipating. Consider co-administration with a stool softener or mild laxative. Efficacy: Prussian blue reduces cesium dose by a factor of two to three by inhibiting enterohepatic recirculation in the GI tract. It is likely to be less effective in inhalation incidents than in ingestion. Clinical Experience: The primary clinical experience with use of Prussian blue was in the Goiânia, Brazil incident in 1987 (IAEA, 1988; 1998a; Maletskos, 1991) and in various cases of thallium poisoning. Experience with Prussian blue in radiation cases has been entirely in ingestion incidents. How Supplied: Prussian blue insoluble capsules contain insoluble ferric hexacyanoferrate (II), with an empirical formula of Fe4 [Fe(CN)6]3. Radiogardase® is supplied as 0.5 g blue powder in gelatin capsules for oral administration. It is packaged in brown glass bottles containing 30 capsules each. The product is manufactured by Haupt Pharma Berlin GmbH for distribution by HEYL Chemischpharmazeutische Fabrik GmbH and Co. KG, Berlin. Prussian blue is stocked in the U.S. Strategic National Stockpile (CDC, 2008). Detailed Treatment with Prussian Blue: Cesium-137 is the dominant cesium radioisotope seen in industrial and laboratory settings. It is also one of the most common radionuclides in aged fission products, and a frequent component of sealed sources used in radiation oncology. Occasionally, other isotopes such as 134Cs are seen in industrial or laboratory settings. Cesium-131 has recently become FDA approved as an implantable seed for treatment of prostate cancer. Thallium acts biochemically like cesium. Radioactive

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thallium is rarely a problem, but poisonings have occasionally occurred with stable thallium. Following systemic uptake, cesium is somewhat uniformly distributed in the body, but the concentration is higher in skeletal muscle than in most other tissues by a few days after intake. The rate of removal from the body depends on age, gender, and other factors (Section 20). In adults, roughly 10 % of the body burden is eliminated over the first few days and nearly all of the remainder is eliminated with an average half-time of ~100 d in males, 75 d in nonpregnant adult females, and 50 d in pregnant females. A tiny portion of the absorbed amount (0.01 to 0.25 %) is eliminated with a half-life on the order of 1 to 2 y. The removal half-life associated with the dominant component of retention increases with age and typically is about two to three weeks in infants, a month at age 5 y, 6 to 7 weeks at age 10 y, and 9 to 11 weeks at age 15 y. Urinary excretion typically accounts for ~85 % of biological removal from the body. Measurement of 137Cs is generally by whole-body counting, or direct gamma counting of biological samples. The most effective means for removing radioactive cesium is the oral administration of ferric ferrocyanide, commonly called Prussian blue. Prussian blue, ferric hexacyanoferrate, Fe4 [Fe(CN)6]3, is an orally-administered drug that enhances excretion of isotopes of cesium and thallium from the body by means of ion exchange. Prussian blue is more effective against ingested than inhaled cesium and thallium. It has had a long and successful history in the treatment of internal depositions of radiocesium and it has been recommended for years as the drug of choice by national and international radiation protection societies. Current FDA guidelines for treatment of internal contamination recommend use of Prussian blue at 3 g tid for a minimum of 30 d (FDA, 2009). Until the Goiânia incident in Brazil in 1987, there were very few reported cases of radiocesium intakes requiring decorporation therapy. However, there is now an increasing potential for such intakes to occur and a need for specific therapy. Prussian blue has a very-high affinity for cesium and thallium, whose metabolism follows an enterohepatic cycle. These ions are ordinarily excreted into the intestine, reabsorbed from the gut into the bile, and then excreted again into the GI tract. Orally-administered Prussian blue traps thallium or cesium in the gut, interrupts its reabsorption from the GI tract, and thereby increases fecal excretion. Thus, the biological half-life of thallium and of cesium are significantly reduced after decorporation therapy with Prussian blue. Prussian blue itself is not absorbed across the gut wall in significant amounts.

206 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL) The efficacy of Prussian blue for treatment of thallium poisoning has been described in various case reports (Atsmon et al., 2000; Barbier, 1974). Even in patients with severe thallium poisoning, most cases published in the literature have had a favorable outcome and the majority of patients with thallotoxicosis have exhibited a good clinical response. Most of the patients who were treated immediately after thallium ingestion with Prussian blue developed minimal traces of thallium-induced neurological symptoms, although the thallium levels were initially high. For some patients, in whom start of treatment was delayed, there was uneventful, although somewhat slower recovery. In some patients with severe clinical signs of thallium intoxication on initial presentation to an emergency department, clinical symptoms were not fully reversible at the time of discharge. Neurological disturbances, particularly in the legs, along with alopecia constituted most of the remaining clinical stigmata. In studies of rats, pigs and dogs internally contaminated with cesium and thallium, the creation of insoluble complexes with Prussian blue in the GI lumen changed the primary elimination route from the kidney to the feces and increased the rate of elimination of these two contaminants. Food is known to increase bile production and enterohepatic circulation and, therefore, additional intake of food may be beneficial in reducing the amount of internal deposition. The increase in enterohepatic circulation may increase the amount of cesium and thallium in the GI lumen, and may increase the amounts available for binding with insoluble Prussian blue. Dose-response studies have not been conducted in humans. In a study using rats (n = 40, mean body weight range of 188 to 219 g) injected with 137Cs, a clear dose-response relationship of the amount of radionuclide eliminated compared to doses of Prussian blue insoluble doses ranging from 1 to 50 mg d–1 was shown. There was little difference in radionuclide elimination rate between the two highest treatment dose of Prussian blue insoluble (50 to 100 mg d–1). In Table 12.10, the “percentage of injected 137Cs remaining” is defined as the percentage of the total injected activity of 137Cs remaining in the body at 96 h post-administration (FDA, 2009). Epidemiological studies and literature review data have been reported in 106 subjects who received Prussian blue after excessive exposure to 137Cs or nonradioactive thallium. Overall, there are literature reports of 65 patients and seven healthy human volunteers who have received Prussian blue after internal deposition of 137Cs. In the 1987 incident in Goiânia, Brazil, 46 persons who were heavily contaminated internally with 137Cs were treated with Prussian blue. Data on the whole-body effective half-life of 137Cs during and

12.3 DECORPORATION THERAPY BY DRUG

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TABLE 12.10—Dose-response data for removal of 137Cs from rats 96 h after administration of Prussian blue.a Insoluble Prussian blue Dose (mg d–1)

Untreated

aFDA

Percentage Injected 137Cs Remaining (range)

58.1

(53.4 – 63.3)

1

9.42 (6.72 – 13.2)

10

1.17 (0.84 – 1.64)

50

0.57 (0.41 – 0.80)

100

0.52 (0.37 – 0.73)

(2009).

after Prussian blue treatment were obtained on 33/46 of these patients. The untreated mean whole-body effective half-life of 137Cs was 80 d in adults, 62 d in adolescents, and 42 d in children. Prussian blue reduced the mean whole-body effective half-life of 137Cs by 69 % in adults, by 46 % in adolescents, and by 43 % in children. Table 12.11 shows the decrease in whole-body effective half-life of 137 Cs in patients during Prussian blue treatment compared with discontinuation of treatment (FDA, 2009). Data from additional literature, including a study of seven human volunteers contaminated with trace doses of 137Cs, and reports on 19 patients contaminated with 137Cs in other incidents, show a similar reduction in whole-body effective half-life after Prussian blue treatment. Thirty-four patients treated with Prussian blue for nonradioactive thallium poisoning are reported in the literature. Prussian blue treatment reduced the mean serum biological half-time of thallium from 8 to 3 d (FDA, 2009). Prussian blue may bind electrolytes found in the GI tract. Asymptomatic hypokalemia, with serum potassium values of 2.5 to 2.9 (normal 3.5 to 5), was reported in 3/42 (7 %) of patients receiving Prussian blue treatment. Serum electrolytes should therefore be closely monitored in patients receiving Prussian blue. Caution should be exercised when treating patients with preexisting cardiac arrhythmias or electrolyte imbalances. Prussian blue may bind some orally-administered therapeutic drugs. As appropriate, blood levels or clinical response to oral medications should be monitored. Reports of adequate and well-controlled drug-drug interaction studies in humans have not been identified in the literature. Binding to some therapeutic drugs and essential nutrients is possible. Well-controlled studies with Prussian blue to evaluate carcinogenesis, mutagenesis and impairment of fertility have not been per-

Age (y)

Prussian Blue (g d–1)

Number of Patients

Adults

>18

10

5

26 ± 6

Adults

>18

6

10

25 ± 15

Adults

>18

3

6

25 ± 9

12 – 14

<10

5

30 ± 12

62 ± 14

4–9

<3

7

24 ± 3

42 ± 4

Group

Adolescents Children

During Prussian Blue Treatment 137Cs T1/2 (d)

Off Prussian Blue Treatment 137Cs T1/2 (d)

80 ± 15 (all 21 adult patients)

208 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL)

TABLE 12.11—Effective half-life of 137Cs during and after treatment with Prussian blue (FDA, 2009).

12.3 DECORPORATION THERAPY BY DRUG

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formed. Studies to determine if Prussian blue is excreted in human milk have not been conducted. Since Prussian blue is not absorbed from the GI tract, its excretion in milk is highly unlikely. However, since cesium and thallium are transmitted from mother to infant in breast milk, women internally contaminated with cesium or thallium should not breast feed. The safety and efficacy of Prussian blue and its dosing for the pediatric population were extrapolated from adult data. Results from pediatric patients internally contaminated with 137Cs and treated with Prussian blue in the Goiânia accident support the use of the extrapolated dosage. Overall, 27 pediatric patients received Prussian blue in the range of 3 to 10 g d–1 in divided doses. Prussian blue treatment reduced the whole-body effective half-life of 137Cs by 46 % in adolescents and by 43 % in children 4 to 12 y of age. In five adolescents and seven children for whom data concerning the rate of radiation elimination are available, the rate was similar to that in adults treated with 3 g tid and in pediatric patients treated with 1 g tid (FDA, 2009). Thallium Contamination: General therapy guidelines for treatment of internal deposition of thallium should follow the radioactive decorporation procedures listed above for 137Cs, except that there is no need for radiation-safety precautions when treating patients contaminated with stable thallium. Patients should also have weekly CBC, serum chemistry, and electrolytes while under treatment. The response to other orally-administered medications should be closely monitored. In cases of severe thallium intoxication, additional types of elimination treatment may be necessary, such as: • induced emesis, followed by gastric intubation and lavage only if initiated within 1 h or so after ingestion; • forced diuresis until urinary thallium excretion is <1 mg 24 h–1; • charcoal hemoperfusion, may be useful during the first 48 h after thallium ingestion (biodistribution phase); • hemodialysis has also been reported to be effective in thallium intoxication; and • multiple dose activated charcoal has also been suggested.

12.3.7 Succimer Treatment [Dimercaptosuccinic acid (DMSA) Chemet®]

210 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL) FDA Indication: Treatment of lead poisoning in pediatric patients with blood lead levels >45 μg dL–1. Radionuclide Decorporation Therapy: FDA approved: lead. Not approved by FDA but considered by others authors: arsenic, bismuth, cadmium, cobalt, mercury, polonium. Notes: 1. Flora et al. (2007) stated that DMSA, D,L-2,3-dimercaptopropane-1-sulfonic acid (DMPS) (not available in the United States), and BAL are the best known treatments against arsenic poisoning, and supplementation of these drugs with antioxidants provide better regimens. 2. Slikkerveer et al. (1998) reported that DMSA and DMPS effectively increase the elimination of bismuth in human urine and may be of benefit in the treatment of patients with bismuth intoxication. 3. Flomenbaum et al. (2006), reported that DMSA may be beneficial for cadmium poisoning with a dose of 10 mg kg –1 three times daily. The same reference recommended treatment for cobalt poisoning treatment is “…GI decontamination and chelation therapy using…EDTA and NAC, and sometimes succimer (DMSA).” 4. According to Rooney (2007), DMPS (not available in the United States) and DMSA are the treatments of choice for mercury toxicity. Modes of Treatment: 100 mg capsules, start at 10 mg kg –1 or 350 mg m–2 orally every 8 h for 5 d, then reduce. Safety and efficacy in pediatric patients <12 months of age have not been established (Table 12.12). Contraindications: Allergy to drug. Pregnancy: Category C (Meadows, 2001). 12.4 Medical Treatments Arranged by Radionuclide 12.4.1 Medical Treatment for Barium and Calcium Radionuclides Radionuclides: Barium (140Ba); calcium (40Ca). Modes of Treatment: For oral ingestion of barium or calcium radionuclides, administer cathartics early. Magnesium or sodium sulfates

Administer

Notes –1

Start dosage at 10 mg kg or 350 mg m–2 orally every 8 h for 5 d. Initiation of therapy at higher doses is not recommended.

Reduce frequency of administration to 10 mg kg –1 or 350 mg m–2 every 12 h (two-thirds of initial daily dosage) for an additional 2 weeks of therapy. A course of treatment lasts 19 d.

Allergy to drug,

DMSA may cause neutropenia or elevate serum transaminase levels. Therefore, CBC and liver functions should be taken before start of treatment and be performed at least weekly during treatment.

As with other chelators, both adults and pediatric patients often experience a rebound in blood lead levels after discontinuation.

No controlled clinical studies have been conducted with succimer in poisoning with other heavy metals.

Repeated courses may be necessary if indicated by weekly monitoring of blood lead concentration.

Patients should be advised to maintain adequate fluid intake and promptly report any sign of infection.

A limited number of patients have received succimer for mercury or arsenic poisoning. These patients showed increased urinary excretion of the heavy metal and varying degrees of symptomatic improvement.

A minimum of 2 weeks between courses is recommended unless blood lead levels indicate the need for more prompt treatment.

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FDA (2009). categories are defined in Appendix J.

bPregnancy

Monitor Clinically

Succimer is generally used as a lead chelator; it forms water soluble chelates and, consequently, increases the urinary excretion of lead.

Pregnancy Category C.b

a

Contradictions

12.4 MEDICAL TREATMENTS ARRANGED BY RADIONUCLIDE

TABLE 12.12—Succimer [dimercaptosuccinic acid (DMSA)] treatment (for arsenic, bismuth, cadmium, lead, mercury).a

212 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL) are the preferred cathartics for barium since they convert barium to insoluble barium sulfate in the GI tract. Consider gastric lavage in the first hour. Elimination of barium and calcium is enhanced with forced diuresis with IV saline and IV furosemide, 1 mg kg –1, to keep urine flow at 3 to 6 mL kg –1 h–1. Serum potassium should be monitored. Calcium radionuclide internal contamination may be treated with calcitonin as for hypercalcemia. Calcitonin reduces serum calcium concentration by increasing renal calcium excretion and by decreasing bone reabsorption. Salmon calcitonin IM or subcutaneously is indicated for treatment of hypercalcemia, while nasal calcitonin is not (Novartis, 2009). For both barium and calcium radionuclides, hemodialysis may also be useful (Leikin and Paloucek, 2008). 12.4.2 Medical Treatment for Cobalt Radionuclides Radionuclides: Cobalt (59Co, 60Co). Summary of Treatment Recommendations: Preferred drug is DTPA (off-label) 1 g IV as described in Section 12.3.3 and in Table 12.13. Alternative drugs to consider, also off-label, are DMSA, EDTA, and NAC at doses described in earlier parts of this section. In mass casualty, administer once. Obtain whole-body counting to evaluate excretion kinetics and to assess for further treatment. Clinical Experience: There is no published clinical experience with the use of DTPA with cobalt inhalation or ingestion. However, there is one case of pediatric cobalt poisoning described in the literature. In that case, EDTA at a dose of 50 mg kg –1 d–1 for 5 d enhanced renal elimination of cobalt, and resolved the patient’s metabolic acidosis and cardiac dysfunction (Flomenbaum et al, 2006). In animal studies, DTPA and EDTA have been shown to be effective in decreasing tissue concentration after intraperitoneal or parenteral injection. N-acetyl cysteine also increases excretion of cobalt in animals and may be considered an alternative (Henge-Napoli et al., 2000). 12.4.3 Medical Treatment for Iodine Radionuclides (FDA, 2001; NAS/NRC, 2004) Radionuclides: Iodine (125I, 131I). Preferred Mode of Treatment: Oral KI is the only form of iodine approved by FDA for use as a thyroid blocking agent, and does not require a prescription. KI is most effective if administered before,

Administer

Ca-DPTA 1 g IV.

Children (<12 y of age) 14 mg kg –1 IV. Administer once unless medically indicated.

Notes

Contraindication

Treatment-induced nephrotoxicity has been noted. Zn-DTPA less toxic.

Monitor Clinically

Obtain whole-body counting or urine and fecal bioassay to evaluate excretion kinetics and to assess for further treatment. There is no clinical experience in the use of DTPA with cobalt inhalation.

12.4 MEDICAL TREATMENTS ARRANGED BY RADIONUCLIDE

TABLE 12.13—Cobalt therapy group (follow Table 12.6; Ca- and Zn-DTPA, off-label for cobalt).

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Predicted Absorbed Dose to the Thyroid [Gy (rad)]b

KI Dose (mg)c

Adults >40 y

≥5 (500)

130

1

2

Adults 18 – 40 y

≥0.1 (10)

130

1

2

Pregnant or lactating women

≥0.05 (5)

130

1

2

Adolescents 12 – 18 yc

≥0.05 (5)

65

0.5

1

1

Children 3 – 12 y

≥0.05 (5)

65

0.5

1

1

1 month – 3 y

≥0.05 (5)

32

0.25

0.5

0.5

Birth – 1 month

≥0.05 (5)

16

0.125

0.25

0.25

Age Category

Number of 130 mg Tablets

Number of 65 mg Tablets

KI Solution 65 mg mL–1 (mL)

a The protective effect of KI lasts ~24 h. For optimal prophylaxis, KI should therefore be administered daily, until a risk of significant exposure to radioiodines by either inhalation or ingestion no longer exists. bWithout KI treatment. c Adolescents approaching adult size (>70 kg) should receive the full adult dose (130 mg).

214 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL)

TABLE 12.14—Threshold thyroid radiation doses and recommended doses of KI for different risk groups (adapted from FDA, 2001).a

12.4 MEDICAL TREATMENTS ARRANGED BY RADIONUCLIDE

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or promptly after, intake of radioactive iodine. KI blocks or reduces the accumulation of radioactive iodine in the thyroid. FDA has approved both tablet and liquid formulations of KI. Summary of FDA Treatment Recommendations (Table 12.14): Individuals intolerant of KI at protective doses, and neonates, pregnant and lactating women (in whom repeat administration of KI raises particular safety issues, see below) should be given priority with regard to other protective measures (i.e., sheltering, evacuation, and control of the food supply). Note that adults >40 y of age need take KI only in the case of a projected large internal radiation dose to the thyroid >5 Gy (500 rad) to prevent hypothyroidism (FDA, 2001). KI should normally be given daily as long as there is continued exposure to radioiodine. For extended length of stay (i.e., >1 d) in contaminated areas, use physician discretion on dose schedule. Alternative Therapy Modes (not FDA approved): • five or six drops of saturated solution of potassium iodide (SSKI) (1 g mL–1) in juice. • Anti-thyroid Agents (propylthioruracil or methimazole): These FDA prescription drugs are indicated for treatment of hyperthyroidism. These drugs block the synthesis of thyroid hormones, so they should be taken early for radioactive iodine contamination before radioiodine organifies. These drugs may be toxic, so the benefit-to-risk should be considered before usage. • Potassiumiodate (KIO3): Not FDA approved in the United States, but utilized in some countries to block thyroid uptake. Painting skin with povidone iodine solution (betadine) has been reported to block the absorption of radioiodine (Moody et al., 1988). Additional KI Treatment Information (Table 12.15): KI should be administered with caution to patients with a history of thyroid disease, iodine hypersensitivity, dermatitis herpetiformis, and hypocomplementemic vasculitis (FDA, 2001). Thyroid disease is much more prevalent in the older population than in infants, children, and adolescents. Administration of KI is intended primarily for mitigation of inhalation dose; control of the food chain is most important for the ingestion pathway. According to IAEA (2005b) and experience from the Chernobyl nuclear reactor accident, the probability of adverse incidents (hypothyroidism, hyperthyroidism, thyrotoxicosis, goiter) is between one

Administer

KI orally, per FDA guidance (Table 12.14)

Additional therapy modes (off-label): SSKIa (1 g mL–1), 5 or 6 drops in juice

Notes

Contraindications

Monitor Clinically

Administration of KI is intended primarily for mitigation of inhaled dose; control of the food chain is most important for the ingestion pathway.

KI should be administered with caution to patients with past or present thyroid disease, iodine hypersensitivity, dermatitis herpetiformis, and hypocomplementemic vasculitis.

According to IAEA (2005b) and experience from the Chernobyl nuclear reactor accident, the probability of adverse incidents (hypothyroidism, hyperthyroidism, thyrotoxicosis, goiter) is 10–6 to 10–7 at the recommended KI dose.

If KI supplies are limited, the highest priority groups to receive stable iodine are newborn infants, lactating mothers, and children, especially <12 y of age.

Thyroid disease is much more prevalent in the older population than in infants, children and adolescents, the target population for treatment with KI.

The risk of death following KI is estimated at 3 × 10–9.

KI should normally be given only once unless there is a continuing release of radioiodine or an extended length of time in a contaminated area.

Neonates especially should be monitored for transient hypothyroidism by measurement of TSH.

216 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL)

TABLE 12.15—Iodine treatment group (for radioactive iodine).

Pregnancy Category Dc Fatal aplastic anemia has been reported. Agranulocytosis is a potential side effect.

Drug hypersensitivity and nursing women.

Monitor complete blood count during treatment.

Propylthiouracild, 50 mg tablets

Pregnancy Category D. Agranulocytosis is a potential side effect.

Drug hypersensitivity and nursing women.

Monitor complete blood count during treatment.

aWebMD

(2009). RxList (2009). cPregnancy categories are defined in Appendix J. dDrugs (2009). b

12.4 MEDICAL TREATMENTS ARRANGED BY RADIONUCLIDE

Methimazoleb (Tapazole®), 5 and 10 mg tablets

/ 217

218 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL) in one million and one in 10 million at the recommended KI dosage. The risk of death following KI is estimated at three in one billion. Neonates especially should be monitored for transient hypothyroidism following KI treatment by measurement of thyroid stimulating hormone (TSH). If KI supplies are limited, the highest priority groups to receive stable iodine are newborn infants, lactating mothers, and children, especially <12 y of age. Efficacy: For optimal protection against inhaled radioiodines, KI should be administered before or immediately coincident with passage of the radioactive cloud, although KI may still have a substantial protective effect even if taken 3 or 4 h after exposure. Use of KI: The dominant initial internal contaminant after a reactor accident, nuclear-weapons test, or any incident involving freshfission products is likely to be 131I. Iodine-131 is also used for therapy in nuclear medicine. Occasionally, other radioisotopes of iodine are encountered. For example, 125I is used for imaging in nuclear medicine. For dosimetric purposes after an accidental exposure, the effective half-life of iodine may be taken to be ~12 d until the specific radionuclides and their relative contributions to the whole are identified. Methods of measurement for iodine isotopes include in vivo thyroid counting and photon spectrometry on biological samples such as urine. The thyroid is the critical organ after intake of radioactive iodine. Traditional thyroid blocking in adults is accomplished by administering 130 mg KI orally as soon as possible after an exposure and one tablet daily for 7 to 14 d. Another convenient way to administer stable iodide is five or six drops of SSKI (1 g mL–1). KI should be administered as soon as possible post-incident, up to 4 h. In situations with continuing exposure, stable iodine may be 50 % effective even 5 to 6 h after contamination with radioiodine. In the immediate vicinity of a nuclear incident (the near field), contamination can begin immediately if the released plume is at a low level. The main route of contamination in the near field is inhalation. A potentially-large thyroid dose might be expected in persons located in the near field. Farther away from the site of the incident (the far field), the main route of contamination with radioiodine would be ingestion of contaminated food and drink, particularly milk. Contamination by these routes could last longer, cover a larger area, and affect a larger population than contamination in the near field. FDA has a variety of information on the use and dosing of KI for the treatment of radioactive iodine (FDA, 2008b). In epidemiological

12.4 MEDICAL TREATMENTS ARRANGED BY RADIONUCLIDE

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studies investigating the relationship between thyroidal radioiodine contamination and risk of thyroid cancer, the estimation of thyroid radiation doses is a critical and complex aspect of the analyses. Estimates of contamination, both for individuals and across populations, have been reached in different studies by the variable combination of: • direct thyroid measurements in a segment of the contaminated population; • measurements of 131I concentrations in the milk consumed by different groups (e.g., communities) and of the quantity of milk consumed; • inference from ground deposition of long-lived radionuclides released coincidentally and presumably in fixed ratios with radioiodines; and • reconstruction of the nature and extent of the actual radiation release. All estimates of individual and population contamination contain some degree of uncertainty. The uncertainty is least for estimates of individual contamination based on direct thyroid measurements. In 1986 the world’s worst nuclear power reactor accident occurred at Chernobyl in the Ukraine (Appendix H). This accident caused massive amounts of radioactive iodine (mainly 131I) and other radioactive products to be spread over parts of the Ukraine, Belorussia and Russia. After the incident, direct measurements of iodine uptake to the thyroid were made in the populations in these three republics of the former Soviet Union (Gavrilin et al., 1999; Likhtarev et al., 1993; Robbins and Schneider, 2000; Zvonova and Balonov, 1993). These thyroid measurements were used to derive, in a direct manner, the thyroid doses received by the individuals from whom the measurements were taken. The thyroid measurements were also used as a guide to estimate the thyroid doses received by other people, taking into account differences in age, milk consumption rates, and ground deposition densities, among other things. These data have a large degree of uncertainty, especially in Belarus, formally Belorussia, where most of the measurements were made by inexperienced people with detectors that were not ideally suited to the task at hand (Gavrilin et al., 1999; UNSCEAR, 2000). Beginning ~4 y after the incident, a sharp increase in the incidence of thyroid cancer among children and adolescents in Belarus and Ukraine was observed. In some regions, observed cases of thyroid cancer among children who were 0 through 4 y of age at the time of the incident exceeded the expected number of cases by 30- to 60-fold. This spike was seen during the first 4 y of the noted increase

220 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL) in thyroid cancer incidence. During the ensuing years, in the most heavily affected areas, the incidence was as much as 100-fold compared to pre-Chernobyl rates (Gavrilin et al., 1999; Likhtarev et al., 1993; Robbins and Schneider, 2000; Zvonova and Balonov, 1993). The majority of cases occurred in children who apparently received <0.3 Gy (30 rad) to the thyroid (Astakhova et al., 1998). A few cases occurred in children exposed to estimated doses of <0.01 Gy (1 rad). However, the uncertainty of these estimates, confounded by medical radiation exposures leaves doubt as to the causal role of these doses of radioiodine (WHO, 1996). Therefore, the best dose-response information from the nuclear reactor accident at Chernobyl shows a marked increase in risk of thyroid cancer in children with exposures of 0.05 Gy (5 rad) or greater (Astakhova et al., 1998; Ivanov et al., 1999; Kazakov et al., 1992). Among children born more than nine months after the accident in areas traversed by the radioactive plume, the incidence of thyroid cancer has not exceeded preaccident rates, consistent with the short half-life of 131I. The use of KI in Poland after the Chernobyl nuclear reactor accident provides us with useful information regarding its safety and tolerability in the general population. Approximately 10.5 million children under 16 y of age and seven million adults received at least one dose of KI. Of note, among newborns receiving single doses of 15 mg KI, 0.37 % (12 of 3,214) showed transient increases in TSH and decreases in free thyroxine (FT4). The side effects among adults and children were generally mild and not clinically significant. Side effects included GI distress, which was reported more frequently in children (up to 2 %, felt to be due to bad taste of SSKI solution) and rash (~1 % in children and adults). Two allergic reactions were observed in adults with known iodine sensitivity (Nauman and Wolff, 1993). Side effects of stable iodine include iodine-induced thyrotoxicosis, which is more common in older people and in iodine deficient areas but usually requires repeated doses of stable iodine. In addition, iodide goiter and hypothyroidism are potential side effects more common in iodine sufficient areas, but they require chronic high doses of stable iodine (Rubery, 1990). Individuals with multinodular goiter, Graves’ disease, and autoimmune thyroiditis should therefore be treated with caution. The transient hypothyroidism observed in 0.37 % (12 of 3,214) of neonates treated with KI in Poland after the Chernobyl nuclear reactor accident has been without reported sequelae to date. There is no question that the benefits of KI treatment to reduce the risk of thyroid cancer outweigh the risks of such treatment in neonates. Nevertheless, in light of

12.4 MEDICAL TREATMENTS ARRANGED BY RADIONUCLIDE

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the potential consequences of even transient hypothyroidism for intellectual development, it is recommended that neonates (within the first month of life) treated with KI be monitored for this effect by measurement of TSH (and FT4, if indicated) and that thyroid hormone therapy be instituted in cases in which hypothyroidism develops (Bongers-Schokking et al., 2000; Calaciura et al., 1995; Fisher, 2000; Mettler et al., 1996). After careful review of the data from the Chernobyl nuclear reactor accident relating estimated thyroid radiation dose and cancer risk in exposed children, FDA revised its recommendation for administration of KI based on age, predicted thyroid exposure, and pregnancy and lactation status (Table 12.14). 12.4.4 Medical Treatment for Radioactive Phosphorus Radionuclide: Phosphorus (32P, 33P). Modes of Treatment: Hydration and oral phosphate binders (one of the following regimens) (Tables 12.16 and 12.17). Calcium carbonate 0.5 to 1 g oral tid (e.g., TUMS® E-X tablet contains 300 mg of calcium). Aluminum hydroxide (600 mg tab; 320 mg 5 mL–1) oral tid. Aluminum carbonate (600 mg tab; 400 mg 5 mL–1) oral tid. Potassium phosphate dibasic, PO 250 mg phosphorus per tablet: • Adult: 1 to 2 tablets PO qid, with full glass of water each time, with meals and at bedtime. • Children over 4 y of age: 1 tab qid. Sevelamer HCL (Renagel®) 400 and 800 mg tablets: • FDA indication: control of serum phosphorus in patients with chronic kidney disease on hemodialysis. • Dose: oral 800 mg tid to 1,600 mg tid. Give as a phosphate binder (use off-label). • Continue therapy × 5 d if possible. The first dose is most important. Obtain urine bioassay to assess treatment efficacy. Pregnancy: Category C (Meadows, 2001). 12.4.5 Medical Treatment for Radium and Strontium Radionuclides Radionuclides: Radium (226Ra), strontium (89Sr, 90Sr).

Sodium glycerophosphate

Route of Administration and Dose

Oral Adult dose, 600 – 1,200 mg phosphate to be given in divided doses.

Sodium phosphate (Na2HPO4)a or potassium phosphate (K2HPO4; Hyperphos-K)b

Oral

Neutral potassium or sodium phosphate (Neutra-Phos K, K-Phos Neutral, Neutra-Phos)b

Oral

Adult dose, 600 – 1,200 mg to be given in divided doses. Adults, 24 mL of one molar dibasic sodium or dibasic potassium phosphate or a mixed neutral salt of monobasic and dibasic sodium phosphate four times daily to supply 3 g of phosphorus in the treatment of hypercalcemia. Pregnancy Category C.c

a

RxList (2009). (2009). cPregnancy categories are defined in Appendix J. bDrugs

Remarks

Not in U.S. Pharmacopeia but in British Pharmacopeia. Well tolerated. Larger doses may result in increased number of soft bowel movements. Chemical grade sodium glycerophosphate has been used without problem. Larger dose (4 g) used as saline cathartic. U.S. recommended daily allowance of phosphorus is 1 g for adults and children over 4 y of age. Sodium salt must be avoided if renal function is inadequate. Potassium phosphate may be used in that case, but not in patients with cardiac insufficiency.

222 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL)

TABLE 12.16—Phosphate drugs. Compound

TABLE 12.17—Phosphorus treatment group. Administer one of the following:

Notes Mechanism of action is dilution and binding. All are oral phosphate binding agents.

Contraindications Use with caution in pre-existing electrolyte abnormalities or in cardiac conditions.

Calcium carbonatea 0.5 –1.0 g PO tid

Monitor electrolytes.

Aluminum hydroxideb [600 mg tab; 320 mg (5 mL)–1] tid Aluminum carbonate [600 mg tab; 400 mg (5 mL)–1] PO tid Sevelamer HCL 800c – 1,600 mg oral tid as a phosphate binder (given off-label)

Binds phosphate in GI tract, therefore probably only effective for first 2 d if 32P still in GI tract. Phosphate drugs may then be used to dilute 32P. Pregnancy Category C.d

Consider parathyroid extract; promotes urinary excretion of calcium and phosphorus. Continue therapy × 5 d if possible. The first dose is most important. aFDA

Pregnancy Category C.d Not approved for pediatric patients.

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(2009) and NLM/NIH (2009). (2009). c RxList (2009). dPregnancy categories found in Appendix J. bDrugs

Monitor Clinically Obtain urine bioassay to assess treatment efficacy.

12.4 MEDICAL TREATMENTS ARRANGED BY RADIONUCLIDE

Administer

Administer a

Aluminum hydroxide [AlternaGEL® (Merck and Company, Whitehouse Station, New Jersey), Amphojel® (American Home Products, Madison New Jersey)] oral. Standard dose for hyperacidity: Adults: 10 mL (1,200 mg) Pediatric: 50 mg kg –1, not to exceed adult dose. Dose to reduce intestinal absorption: 60 to 100 mL (NCRP, 1980)

Notes Post-intake action, these regimens block intestinal absorption. Also inhibit absorption by competitive action. In mass casualty, administer once for a given regimen.

Barium sulfat,eb 100 to 300 g oral in a single dose in 250 mL water. Alternative regimens: Sodium alginate,c 5 g oral twice daily, then 1 g four times daily with water.

Alginate is an ingredient in Gaviscon® (GlaxoSmithKline, London), ~200 mg per tablespoon or tablet. Do not exceed use as directed on label.

Contraindications Hypersensitivity

Monitor Clinically Can cause constipation. Prolonged use of aluminum hydroxide may cause hypophosphatemia. Since radium and strontium are bone seekers, blood counts should be monitored for pancytopenia.

224 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL)

TABLE 12.18—Drugs to block absorption of radium or strontium [doses are for adults unless otherwise specified (Leikin and Paloucek, 2008; NCRP, 1980)].

aDrugs

Calcium increases urinary excretion of strontium, and phosphate decreases intestinal absorption of strontium (NCRP, 1980).

(2009) and FDA (2009). Drugs (2009), Mayo (2009), and NLM/NIH (2009). cNutrition Surplus (2009). dDrugs (2009). b

12.4 MEDICAL TREATMENTS ARRANGED BY RADIONUCLIDE

Calcium phosphate,d 1,200 mg oral once.

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226 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL) Modes of Treatment: One or more of the drugs in Table 12.18 should be taken as soon as possible post-incident to block intestinal absorption. Alternate Regimens: Alginate, a nonprescription ion-exchange resin that works in the GI tract, may bind strontium and other radionuclides. (Alginate is an inactive ingredient in Gaviscon,® ~200 mg per tablet or per tablespoon. Do not exceed dose per label.) Try alginate dose 5 g twice daily for 1 d then 1 g qid (Bhattacharyya et al., 1992). Calcium phosphate; 10 g oral × 1. After Absorption: Calcium gluconate 1 g IV slowly over 5 to 15 min (for strontium) (Table 12.19). Diluting Agents: Strontium lactate 500 to 1,500 mg d–1. Strontium gluconate 600 mg daily IV for up to 6 d. In a mass casualty incident, administer once for each regimen. Contraindications: Too rapid an IV infusion of calcium gluconate may precipitate hypotension. Treatment may also lead to constipation. Strontium Biokinetics: Strontium-90 is the predominant isotope of interest, although other strontium isotopes are occasionally encountered. In particular, 89Sr has been used successfully for palliative therapy in bone pain secondary to metastatic cancer. A comprehensive age-specific model for strontium retention has been developed (Section 20). The model depicts transfer to soft tissue, cortical and trabecular bone volume and surface, two liver compartments, and the renal system. The transfer rates in this model to cortical and trabecular bone are dominant. For most forms of 89Sr or 90Sr, doses to bone surface and red marrow are the main concern following intake by either inhalation or ingestion; inhaled strontium titanate is an exception due to its prolonged retention in the lung and relatively-high dose to that organ. Because of the biologically significant rate of strontium transfer to the GI tract, it is also necessary to block intestinal absorption in those cases where intake is by inhalation. The following treatments are useful in the medical management of inhalation cases with strontium: • IV calcium gluconate 2 g in 500 cc over 4 to 6 h (competes for bone binding sites); and • barium sulfate 300 g orally as soon as possible post-incident to block intestinal absorption.

Route of Administration and Dose

Compound

Calcium carbonate [e.g., Titralac® (3M, St. Paul, Minnesota) and TUMS® (GlaxoSmithKline, London)]a

Oral

Calcium gluconateb ampules

Intravenous

Remarks

May cause constipation.

As directed on label.

5 ampules each containing 500 mg calcium can be given in 500 mL 5 % glucose in water over 4 h (NCRP, 1980).

Can be given daily on six consecutive days. IV calcium should not be given to persons receiving quinidine or digitalis preparations or to those who have a very slow heart rate (NCRP, 1980). Too rapid an IV infusion of calcium gluconate may precipitate hypotension. Monitor blood pressure during administration.

aDrugs bRxList

(2009). (2009).

12.4 MEDICAL TREATMENTS ARRANGED BY RADIONUCLIDE

TABLE 12.19—Calcium compounds after absorption of strontium (do not use for chromium decorporation).

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228 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL) 12.4.6 Medical Treatment for Tritium (radioactive hydrogen) Radionuclide: Hydrogen (3H). Modes of Treatment: Force liquids to 3 to 4 L d–1. May be increased to 6 to 10 L d–1 in severe internal deposition if not clinically contraindicated; continue × 5 d (Tables 12.20 and 12.21). Increasing liquids to a maximum level of tolerance will decrease the biological half-life of 3H by one-half to one-eighth of its normal value, thereby reducing the dose proportionally. Contraindications: Caution should be used in patients in whom there is concern for fluid overload, such as those with congestive heart failure. Tritium Biokinetics: Tritium is the only radioactive isotope of hydrogen, decaying to 3He with emission of an electron with maximum energy of 18.6 keV. Tritiated water (HTO) is taken easily into the body by inhalation, ingestion, or by transdermal absorption. Uptake of 3H in the body is complete and it is assumed to be instantaneously absorbed and mixed with body water. Dose to total-body water is therefore the critical issue in the management of internal contamination with tritium. A biokinetic model for inhaled or ingested HTO is described in Section 20. Medical management of tritium intake is primarily directed at increasing body water turnover. Single contamination episodes are treated by increasing oral fluid intake. This has the dual value of diluting the tritium and increasing excretion by physiological mechanisms. An increase in oral fluids of 3 to 4 L d–1 reduces the biological half-life of tritium by a factor of two to three and therefore reduces whole-body dose in the same proportion. Bioassay is generally accomplished by analyzing 24 h urine collections with liquid scintillation counting. For larger contaminations, intravenous hydration, management of fluid intake and output, and use of diuretics is a possible modality for increasing turnover of body water, but, historically, this has rarely been necessary. 12.4.7 Medical Treatment for Uranium Isotopes Radionuclides: tion 20).

234

U,

235

U,

238

U, and depleted uranium (DU) (Sec-

Modes of Treatment (Table 12.22): All drugs listed are by prescription, off-label, for adults: Isotonic sodium bicarbonate, 250 mL (1 to 2 mEq kg –1) slow IV infusion, or

TABLE 12.20—Tritium treatment group (NCRP, 1980). Notes

Contraindications

Force liquids to 3 to 4 L d–1. May be increased to 6 to 10 L d–1 in severe contaminations if not clinically contraindicated.

Increasing liquids to maximum level of tolerance will decrease the biological half-life of 3H by 0.5 to 0.125 of its normal value, thereby reducing dose in the same amount.

Use with caution in cardiac or renal disease patients who could be susceptible to fluid overload.

Continue × 5 d.

Monitor Clinically

Fluid and electrolytes

TABLE 12.21—Forced fluids (NCRP, 1980). Compound

Route of Administration and Dose

Remarks

Oral, 3 – 4 L d–1

Usually forced to the tolerance level by patient. Check urine volumes and save voidings separately for radioassay.

5 % glucose in water or saline

Intravenous, Up to 3 L d–1

Used only if fluids cannot be given orally. Usual precautions on electrolyte balance (sodium and potassium) must be observed.

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Water (or other liquids such as soda, juice, tea, coffee, beer)

12.4 MEDICAL TREATMENTS ARRANGED BY RADIONUCLIDE

Administer

Administer

Notes

Isotonic sodium bicarbonat,ea 250 mL slow IV infusion, or two bicarbonate tablets oral q4h until the urine reaches a pH of 8 – 9. Continue × 3 d.

Alkalinization of the urine promotes excretion of U-carbonates, reduces probability of acute tubular necrosis. Alkalosis has occurred after one tablespoonful in a young infant. Pregnancy Category C.b

For high level of uranium intake, consider off-label diuretic dosing as follows: Etidronate (Didronel)a – 400 mg PO daily or

Pregnancy Category C.

Diamoxa – 500 mg PO bid. Continue × 3 d.

Pregnancy Category C.

Consider dialysis for suspected high levels of intake. aDrugs

(2009). categories are defined in Appendix J.

bPregnancy

Contraindications

Anuria

Monitor Clinically

Obtain urine pH hourly. Blood pH, serum electrolytes and renal functions should be monitored.

A preexisting hypokalemia may be unmasked and this treatment should be used with caution in congestive failure or in disease states with sodium retention.

230 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL)

TABLE 12.22—Uranium treatment group.

12.4 MEDICAL TREATMENTS ARRANGED BY RADIONUCLIDE

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two bicarbonate tablets PO q4h until the urine reaches a pH of eight to nine. Monitor urine pH hourly (Thomson MICROMEDEX). Continue × 3 d. For suspected high level of uranium intake, consider dosing as follows: Consider renal dialysis. Etidronate (Didronel) – 400 mg PO daily or Diamox – 500 mg PO bid Continue × 3 d. Precautions: Research in animals led Stradling and Taylor (2005) to conclude that DTPA is not an effective treatment for soluble uranium, thorium or neptunium. 2+ Alkalinization of the urine promotes excretion of UO 2 - complexes, and reduces the likelihood of patients developing acute tubular necrosis. Urine pH and serum electrolytes and renal functions should be monitored. A preexisting hypokalemia may be unmasked with the use of these therapies and this treatment should be used with caution in patients with congestive heart failure or in those having disease states with sodium retention. Precaution: There is controversy to the use of sodium bicarbonate. Stradling and Taylor (2005) states, “The evidence available suggests it (sodium bicarbonate) is not effective, and may cause hypokalemia and respiratory acidosis.” Uranium Biokinetics: The biological behavior of uranium in the human body is discussed in Section 20. The most common uranium isotopes seen in research and in industry are 238U, 235U, and 234U. Uranium compounds can exist in various solubility classes as noted below: UF6: uranium hexafluoride UO2(NO3)2: uranyl nitrate UO2: uranium dioxide UO2: high-fired uranium dioxide UO3: uranium trioxide U3O8: uranium oxide

Inhalation Type: F Inhalation Type: M Inhalation Type: M (may initially be S) Inhalation Type: S Inhalation Type: S Inhalation Type:S

232 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL) Inhalation is the usual route of internal contamination in the workplace. The acute toxicity of uranium is due to its chemical rather than its radiological properties, particularly with regard to the renal system. It is assumed that all excretion occurs via the urine. Except possibly for intake of highly-enriched uranium, chemical toxicity is expected to be the predominant mechanism for renal damage. (Natural uranium has 0.7 % 235U by weight and DU has ~0.2 to 0.3 % by weight of 235U.) The renal toxic properties are therefore the basis of occupational exposure limits. In acidic urine, the uranyl ion binds to renal tubular surface pro2+ teins, and some of the bound UO 2 is retained in the kidney. The kidney is the first organ to show chemical damage in the form of nephritis and proteinuria. In an extensive review of both human and animal data, Kathren and Burklin (2008) concluded that the threshold for permanent kidney damage “apparently incurred without significant or long lasting ill effects” is 3 µg U g–1 kidney tissue (~0.013 mg kg–1 body weight or ~1 mg in a 70 kg person) with a “likely safety factor of 10 to 100.” In an analysis of 27 cases of individuals exposed to uranium, Roszell et al. (2009) defined four renal effects groups. No detectable effects are likely in individuals with kidney uranium concentrations ≤2.2 µg g–1 kidney tissue (~0.01 mg kg–1 body weight). At uranium concentrations >2.2 to ≤6.4 µg g–1 (>0.01 to ≤0.03 mg kg–1 body weight) transient renal dysfunction is possible, but not severe enough to cause illness. At uranium concentrations >6.4 to ≤18 µg g–1 (>0.03 to ≤0.08 mg kg–1 body weight) protracted indicators of renal dysfunction and illness are possible. At uranium concentrations >18 µg g–1 kidney tissue (>0.08 mg kg–1 body weight), severe clinical symptoms of renal dysfunction are possible and illness is likely. This would be equivalent to ~6 mg in a 70 kg person. The current occupational maximum permissible concentration of uranium is 3 µg g–1 kidney (~1 mg in a 70 kg person) (Roszell et al., 2009), which may not be sufficiently protective (Leggett, 1989). Evidence of permanent renal damage is seen with a permanent increase in bound urinary nitrogen and creatinine, along with proteinuria and a decrease in glomerular filtration rate. Kathren and Burklin (2008) conclude that for soluble uranium compounds the 50 % lethality level exceeds several grams for oral intakes and is at least 0.41 g for inhalation intakes. For humans they suggest 5 g uranium as the “provisional” acute oral LD50 and 1 g as the “provisional” acute inhalation LD50. If the kidneys are contaminated with high levels of uranium for prolonged periods of time, the original tubular epithelial cells are replaced by epithelial cells which appear to be more resistant to uranium (Kumar et al., 2005) (see Section 20.24 for additional information on uranium).

12.5 LUNG LAVAGE

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Oral doses or infusions of sodium bicarbonate are the treatment of choice in uranium contamination. Treatments should be titrated to keep the urine alkaline and monitored by frequent pH measurements. The nontoxic uranium carbonate complexes are increased by three to four orders of magnitude in alkaline urine and efficiently excreted. ICRP (1978) recommends use of a diuretic drug. DTPA at a dose of 100 mg kg –1 increased the LD50 in mice only if given within 4 h of the contamination (Catsch and HarmuthHoene, 1979). Use of DTPA for uranium decorporation would be 2+ off-label and is only effective in large UO 2 intakes. Depleted Uranium: Depleted uranium (DU) is a byproduct of the uranium enrichment process and has a lower percentage of 234U and 235U than natural uranium. The activity of DU is ~60 % of the activity of natural uranium so the main concern with DU is its chemical toxicity as a heavy metal. U.S. military veterans with DU exposure and contamination have been followed by the Baltimore Veterans Affairs DU Follow-Up Program since 1993. Soldiers with embedded DU fragments have continued to excrete elevated levels of uranium in their urine but there have been no clinicallysignificant uranium-related health effects identified to date. Surveillance of these soldiers will continue (NAS/NRC, 2008a; Squibb and McDiarmid, 2006). 12.4.8 Medical Treatment for the Actinide Nuclides (Section 12.3.3) The actinides are the heavy elements at the bottom of the Periodic Table of the Elements beginning with actinium (Z = 89) and ending with lawrencium (103). They include actinium, thorium (90), protactinium (91), uranium (92), neptunium (93), plutonium (94), americium (95), curium (96), berkelium (97), californium (98), einsteinium (99), fermium (100), mendelevium (101), nobelium (102), and lawrencium. Medical Management: Ca- and Zn-DTPA chelation therapy is the treatment of choice for all routes of internal contamination involving actinides except for thorium, uranium and neptunium (Stradling and Taylor, 2005) (see management with DTPA in Section 12.3.3). Ca-EDTA may be used if DTPA is not immediately available, but it is substantially less effective and more toxic. 12.5 Lung Lavage Lavage of the tracheobronchial tree (bronchoalveolar or lung lavage) is a technique to consider in the treatment of individuals

234 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL) who have inhaled relatively-large amounts of insoluble radionuclides. According to Stradling and Taylor (2005), lung lavage should only be undertaken if there is a likelihood of deterministic effects. For example, lung lavage should only be considered if the estimated lung dose is likely to exceed 5 Sv (500 rem) within a few weeks. Lung lavage has been used to remove inflammatory exudates and other foreign materials from the lungs of humans suffering from a variety of chronic obstructive lung diseases. The procedure requires a general anesthetic. The entire volume of one lung at a time is filled with isotonic saline, or other irrigation fluid. The lung is then filled and drained repeatedly by gravity flow. At the end of the procedure as much fluid as possible is removed from the treated lung by drainage followed by suction. Lung lavage performed by an experienced pulmonologist and anesthesiologist is a fairly safe procedure, especially since the procedure has been improved to allow lavage of one lung lobe at a time. The use of lung lavage for removal of inhaled radioactive aerosols has been studied experimentally in dogs and baboons. Dogs were lavaged after inhalation of 144Ce, 137Cs, or 95Zr / 95Nb contained within insoluble fused aluminosilicate particles (FAP) (Boecker et al., 1974; Muggenburg et al., 1975; Pfleger et al., 1969). A single lavage of one lung will remove ~12 % of the initial lung burden, while lavage of both lungs will remove about twice that value (Muggenburg et al., 1977). An initial lung burden may be reduced by ~25 to 50 % (average was 44 % in eight dogs) by means of 10 saline lavage treatments, five for each lung. Each lung was lavaged once each week during the first two weeks after contamination followed by lavage of alternate lungs each week out to 56 d. A 53 % reduction of cumulative radiation dose to the lung resulted from these treatments; radiation pneumonitis and early deaths were prevented in the majority (75 %) of the treated dogs in contrast to the untreated dogs (Muggenburg et al., 1975). In experiments with baboons inhaling plutonium oxide, treatment with 10 pulmonary lavages resulted in removal of 60 to 90 % of total lung burdens (Nolibe et al., 1976). Additional experiments in dogs indicated that 10 lavage treatments within 24 d gave approximately the same results as 10 lavages in 56 d (Silbaugh et al., 1975). Ten more lavages (20 total) removed an average of 52 % of the initial lung burden, a total amount that was not greatly different from results of the original 10 lavage schedule. Studies on 144Ce in insoluble FAPs indicate that a sizable percentage (>80 %) of the insoluble particles in the lung continued to be accessible to removal by bronchopulmonary

12.5 LUNG LAVAGE

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lavage for periods up to six months after contamination, but translocation from the lungs to other tissues does increase with time (Felicetti et al., 1975). Thus, although early treatment is desirable when the deposited radionuclide(s) has a high dose rate, the time after contamination that lung lavage treatments are initiated may not be critical, up to several months after exposure, for the removal of material with a long lung retention time (Muggenburg et al., 1977). In one human case of an inhaled 239Pu aerosol, three lavages (8, 12, and 17 d post-contamination) removed ~13 % of the estimated initial lung burden (McClellan et al., 1972a). The aerosol proved more soluble than initially assumed and this factor may have reduced the effectiveness of lavage therapy. The excretion of 239Pu in the urine following daily intravenous administration of CaDTPA, starting on day eight after contamination, was ~17 % of the initial lung burden. Muggenburg et al. (1977) point out that lavage and chelation therapy can be used effectively together for heterogeneous aerosols or aerosols of unknown solubility. Possible use of this technique in humans requires a careful benefit-to-risk assessment. The risk of the procedure lies with both the administration of a general anesthetic and any mortality or morbidity that could be associated with the procedure. Muggenburg et al. (1972) found that dogs serially sacrificed after lavage had no physiologic or pathologic changes one week after the procedure. Early physiologic alterations resolved within 24 h and only vestiges of tissue reaction were found by light microscopy at 48 h. After three lavages of the individual contaminated with 239Pu, there was neither evidence of impaired lung function nor signs of adverse reactions 24 h after the last procedure. Studies of surfactant lipids in the lungs of dogs show that the lipids removed by lavage are replaced in ~5 h (Henderson et al., 1975). Muggenburg et al. (1976a) have shown in dogs that multiple lung lavages carry little biomedical risk and that the primary risk is associated with general anesthesia. In a review of clinical and experimental uses of bronchopulmonary lavage, Muggenburg and Jones (1971) noted that a single death was recorded after 204 lavages in 70 patients who had various types of severe obstructive lung disease. In dogs, two deaths were recorded after 420 lavages in 149 animals. An update on humans cites one death after 258 lavages in 112 persons (Muggenburg et al., 1976b). Nolibe et al. (1976) report five deaths in baboons in the course of 800 lavages. This overall experience suggests a possible mortality rate of ~0.5 % for each lavage procedure. Since the human cases were complicated by the presence of debilitating

236 / 12. STAGE 7: MEDICAL MANAGEMENT (HOSPITAL) pulmonary disorders, the estimated risk is probably greater than is to be expected in otherwise healthy individuals. Because lavage requires general anesthesia, the risks should be comparable with the general experience with low-risk surgical procedures. Deliberations on the justification of the procedure should consider the age and health of the patient relative to the possible deleterious effects of the radiation exposure. Contamination that carries a high risk of early radiation effects should be weighed against the risk of intervention. In general, though, the potential effects from internal contamination will occur 10, 20, or more years in the future while the risk of lavage is immediate. Elderly or chronically ill individuals may, therefore, experience relatively little benefit from such interventional procedures. In addition to its therapeutic role, bronchopulmonary lavage may help with dose determination. Assay of radionuclides in recovered lavage fluid can be used to estimate the lung burden of radioactive material. Guilmette et al. (1986) have developed a standard bioassay procedure for cases of plutonium inhalation.

Part D: Patient Management Post-Hospital 13. Stage 8: Follow-Up Medical Care

Objectives • observe for: - possible late deterministic effects, - latent acute radiation syndrome, - psychosocial distress symptoms, and - radiation related cancers. • monitor internal contamination levels, assess need for further treatment; and • consider for contribution to epidemiological studies. See Figure 3.1 for flow of persons through all nine stages in the management of radionuclide contamination. 13.1 Introduction In the continuing medical evaluation of internal exposure cases, it is common to distinguish late physiological and psychological effects from the early effects of radiation exposure. However, when compared with high-level external exposure, medically-significant organ pathology resulting from internal exposure is relatively rare. Most deterministic effects occur early and typically show a sigmoid dose-response curve above an appropriate threshold, and the severity of harm from the radiation exposure increases with dose. Effects are non-neoplastic and are rather promptly expressed 237

238 / 13. STAGE 8: FOLLOW-UP MEDICAL CARE in the exposed individual. In contrast, late stochastic effects represent a probabilistic tissue response to radiation exposure. Stochastic effects are nonthreshold and are generally expressed later in an exposed population. Follow-up medical care of an internally-irradiated individual will therefore focus on the evaluation of late effects, most significantly directed to the detection of cancer. In addition, late psychological issues related to radiation exposure have been shown to be quite important and these issues should always be considered by the examining physician as part of continuing medical surveillance. In the case of a relatively-low-dose internal exposure, the long-term psychological trauma may be more medically significant than any radiation-induced organ damage. 13.2 Late-Occurring Health Effects Radiation-induced damage in reproductive cells may give rise to inheritable genetic mutations, while damage in somatic cells may increase the chance of neoplasia. Genetic mutations from ionizing radiation have long been shown in animal studies, but current evidence in humans is inconclusive, likely due to a relatively-high baseline of naturally-occurring genetic lesions. In studies of radiation-induced carcinoma, leukemias and bone cancers have shown a minimum latency time of 2 to 5 y, while solid tumors have a minimum latency time of 10 y or more. For solid tumors, excess tumors commonly occur at ages comparable to those at which they occur naturally. For carcinomas of normal adult onset, current evidence suggests that the latency period is >10 y and tumors may arise at their normal adult age. From the medical effects seen in survivors of the Japanese atomic bombings, which were predominantly due to relativelyhigh-level external neutron and gamma dose delivered at a high dose rate, there is statistically significant evidence for all radiation-induced leukemias (except chronic lymphocytic leukemia), breast, thyroid, colon, stomach, lung, ovarian carcinoma, and borderline or inconsistent results for radiation-induced carcinoma of the esophagus, liver, skin, bladder, central nervous system, multiple myeloma, and lymphoma. Regarding noncancer disease, there is also strong evidence for radiation-induced cataracts, hyperparathyroidism, a decrease in the T-cell mediated and humoral immune response, and various types of chromosomal aberrations in lymphocytes and other cell lines (ICRP, 2007). An International Agency for Research on Cancer Study Group evaluated combined mortality data for 96,000 nuclear industry workers in the United States, Canada, and the United Kingdom

13.3 PREVENTIVE MEDICINE APPROACHES

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(IARC, 1994). The exposure was primarily to low-level gamma radiation and the risk analysis was based on a constant linear excess relative risk (ERR) model in which ERR = 1 + b(dose), where b is a constant. In this study, ERR was found to be 2.2 Sv–1 (220 rem) for leukemia, while ERR for all cancers, excluding leukemia, was essentially zero. Additional risk estimates from large epidemiological studies in Europe show similar results (Cardis et al., 2005). 13.3 Preventive Medicine Approaches For adult patients with internal exposure from an occupational radiation incident or from a terrorism attack, the primary preventive medicine emphasis should be on cancer surveillance using currently available and accepted medical screening tools appropriate for the patient’s age. If children or young adults are exposed to radiation from internally-deposited radionuclides, they should also be routinely screened periodically for cancer. The U.S. Preventive Services Task Force generally issues periodically-updated recommendations for age-dependent tests for cancer and for other chronic diseases. This Task Force, first convened by the U.S. Public Health Service in 1984, and since 1998 sponsored by the Agency for Healthcare Research and Quality, is the leading independent panel of private-sector experts in prevention and primary care (AHRQ, 2008). This Panel conducts rigorous, impartial assessments of the scientific evidence for the effectiveness of a broad range of clinical preventive services, including screening, counseling, and preventive medicine. Its recommendations are considered the primary standard for population-based clinical preventive services and should generally be followed in evaluating patients with prior radiation exposure. However, in the medical care of an individual patient in the private sector, the final analysis of what tests are appropriate rests with physician judgment since the physician has detailed knowledge of the patient’s total medical history, including radiation exposure. Patients who have undergone therapy for radionuclide intakes should be considered for long-term follow-up bioassay consistent with the radionuclide of exposure. The nature and extent of this follow-up is case-specific, depending on the radionuclide, mode and magnitude of intake, and nature of the therapy. This bioassay would most likely consist of urine sample analysis or in vivo measurements. The purpose of the long-term follow-up is threefold. First, it provides data to verify the accuracy of intake and dose estimates which can indicate the effectiveness of therapy. Second, it establishes a baseline against which future measurements can be used to identify the possibility of a new intake occurring. This latter point

240 / 13. STAGE 8: FOLLOW-UP MEDICAL CARE is particularly important for occupational workers who have undergone therapy and subsequently returned to work with the same radionuclide for which they underwent treatment. Third, it provides data for a future dose assessment if such is needed whether for the benefit of the individual or for an epidemiology study. In the U.S. government sector, the DOE Office of Occupational Medicine and Medical Surveillance has supported the Former Radiation Worker Medical Surveillance Program at the various DOE facilities as part of their occupational and preventive medicine outreach program (DOE, 2008b). It is expected that participation in this program will be entirely voluntary and provide the exposed patient with long-term medical monitoring and a periodic update of the assessment of their radiation dose. It is also expected that participating patients will receive periodic medical examinations and in vivo and in vitro bioassay measurements of residual radioactivity as appropriate. Generally, the causation of many of the higher internal doses would be incidents that are well documented. Informed consent should be documented using a consent form approved by an appropriate institutional review board. It is also desirable that demographic, medical and dosimetric information of exposed workers be maintained in an easily available computer database and be evaluated for any trends or correlations between exposure and health outcome. Medical surveillance from radiation incidents in the nongovernment sector would be well advised to consider this U.S. government model for patient surveillance. REAC/TS is the deployable medical asset of the DOE complex and is available to provide acute and long-term medical evaluation of radiation incidents, both in the private and federal government sectors. REAC/TS has functioned as a DOE asset since 1976 and has provided medical evaluation and advice in most recent significant instances of either internal or external radiation exposure (Appendix G). 13.4 Psychosocial Issues Various aspects of the psychological issues that will likely need to be addressed by the medical community after a radiation incident have been discussed by the U.S. Department of Homeland Security Medical Working Group on RDD Preparedness (DHS, 2003) and various researchers including Becker (2001; 2004; 2005; 2007), Becker and Middleton (2008), Bromet et al. (2000), Collins (1992), Ginzburg (1993), Pastel et al. (2001), Ursano (2002), Vyner (1983; 1988), and Yehuda (2002). These issues would involve psychological support and assistance in the acute phase of an incident

13.4 PSYCHOSOCIAL ISSUES

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as well as long-term follow-up activities. A terrorist attack involving the release of radiation will create uncertainty, fear and terror in the general population. Following the detonation of an RDD, an improvised nuclear device (IND), or any other incident involving either external irradiation or internal exposure, the management of acute psychological and behavioral responses is likely to be as important and challenging as the treatment of trauma or radiation-related medical injuries and illnesses. Furthermore, the psychological effects of radiation exposure can manifest years after the causative exposure and the patient may experience fear for the safety of future generations. Those who have been exposed may also experience feelings of vulnerability, chronic anxiety, and lack of control. For internal dosimetry calculations, it is also common for some time period to elapse before the bioassay samples have been analyzed and a calculated intake and internal dose assigned. During this time period, a lack of consensus among experts can also increase patient fear and anger. Affected individuals appear to fall into one of three groups: those who are distressed, those who manifest behavioral changes, and those who may be at high risk to develop psychiatric illness. Distress may be common and manifest as sadness, anger, fear, difficulty sleeping, impaired ability to concentrate, and disbelief. Psychological distress after a radiological incident may also manifest as nonspecific somatic complaints. This condition is often referred to as multiple idiopathic physical symptoms (MIPS). Engel et al. (2002) evaluated various aspects of the MIPS issue. Competing-risks analysis was used to determine the 1 y longitudinal outcomes, including mortality, associated with MIPS in a population sample. The authors analyzed baseline and 1 y follow-up data from a population-based National Institute of Mental Health Catchment Area Epidemiological Study. This study was initiated in response to the 1977 report of the President’s Commission on Mental Health (PCMH, 1978). The purpose was to collect data on the prevalence and incidence of mental disorders and on the use of and need for services by the mentally ill. In the study by Engel et al. (2002), multinomial logit regression was used to examine the incidence of MIPS, resolution of such symptoms, and related mortality among individuals in the general population, with adjustment for demographic characteristics and the presence or absence at baseline of a lifetime diagnosis of major depression, anxiety disorder, and alcohol abuse. Most of the individuals with MIPS recovered over the ensuing year. The incidence of MIPS among those without such symptoms at baseline was 1.7 %. The predicted mortality among individuals

242 / 13. STAGE 8: FOLLOW-UP MEDICAL CARE with MIPS at baseline was higher than for individuals not having such symptoms at baseline (0.28 versus 0.18 %). The higher mortality rate among those with MIPS at baseline persisted after adjustment for covariates and competing outcomes. Outcomes associated with MIPS vary widely. Most individuals improve over time. However, the course for a few individuals is less benign than perhaps previously thought. Further research is needed to determine the mechanisms behind increases in mortality related to MIPS, the predictors of poor prognosis, and whether mortality remains elevated over longer periods of follow-up. When properly informed on these potential long-term effects, general health-care providers, working in conjunction with mental health services, should be able to care for these patients. Parents who have experienced radiation exposure, either internal or external, have been noted to exhibit behavioral changes such as decreasing their travel, staying home, refusing to send children to school, and increasing substance use and abuse. In the case of a mass casualty incident, children may also be particularly vulnerable to psychological issues (Bromet et al., 2000; Korol et al., 1999; Weinberg et al., 1995). Stigmatization of those exposed or traveling from contaminated areas can be expected. This will affect the relocation and entry of new students into school systems. Outreach offering of health education to school systems, parent-teacher education programs, and school-nurse training can allay community anxiety. For some individuals, symptoms may persist, affecting function at home and work, and possibly even resulting in overt psychiatric illness. However, for the majority of people, distress and psychological and behavioral symptoms related to the traumatic incident exposure are likely to diminish over time. It should also be borne in mind that first responders and first receivers are also likely to have major concerns and information needs during and after a radiation incident. It will be important for these to be addressed. At the request of CDC, researchers at the University of Alabama at Birmingham (Becker and Middleton, 2008) conducted a series of 10 focus groups (77 total participants) with emergency department physicians and nurses at hospitals in three U.S. regions. Hospital emergency department clinicians will play a crucial role in responding to the acute issues pursuant to a terrorist incident involving radioactive materials. To date, however, there has been a paucity of research focusing specifically on emergency department clinicians’ perspectives regarding this threat. Participants in the focus groups considered a hypothetical “dirty bomb” scenario and discussed their perceptions, concerns,

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information needs, preferred information sources, and views of current guidance and informational materials. Study participants consistently expressed the view that neither emergency departments nor hospital facilities are sufficiently prepared for a terrorist event involving radioactive materials. Key clinician concerns included the possibility of the hospital being overwhelmed, safety of loved ones, potential staffing problems, readiness problems, and contamination and self protection. Participants also expressed a need for additional information, strongly disagreed with aspects of current response guidance, and in some cases indicated they would not perform current protocols. This study is the first to examine the views, perceptions, and information needs of hospital emergency department clinicians regarding radiological terrorism. As such, the findings may be useful in informing current and future efforts to improve hospital preparedness.

14. Stage 9: Contaminated Decedents (hospital and mortuary)

Objectives • protect medical and mortuary professionals from unnecessary radiation exposures; • control radionuclide contamination of individuals and facilities; and • ensure proper disposal of decedent. See Figure 3.1 for flow of persons through all nine stages in the management of radionuclide contamination. 14.1 Introduction In spite of the best efforts of medical caregivers, patients’ injuries may prove fatal. Some of these patients may be contaminated, perhaps heavily, and they will need to be treated not only with appropriate respect, but also to minimize the spread of contamination. Some guidelines are given in ICRP Publications 94, 96, and 98 (ICRP, 2004; 2005a; 2005b); NCRP Report No. 37 (NCRP, 1970), and Wood et al. (2007; 2008). However, NCRP Report No. 37 and ICRP Publication 94 address patients administered relativelylarge doses of radiopharmaceuticals, and ICRP Publication 98 discusses patients who have been implanted with a large number of sealed radioactive sources, not patients exposed to relatively-low amounts of activity as contamination.9 The most recent and complete guidance on this topic is a CDC publication entitled Guidelines for Handling Decedents Contaminated with Radioactive 244

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Materials (Wood et al., 2007). Accordingly, the chief concern should be the control of contamination and not reducing risk to medical examiner and mortuary personnel. However, it is still wise for all persons involved in transportation, preparation, and final disposition of contaminated decedents to take appropriate precautions, which will be discussed. Persons involved in radiological incidents are likely to be contaminated, perhaps heavily, and their bodies may remain contaminated after their death. Although dose rates from even relatively-high levels of contamination are not likely to be high,10 certain precautions are still recommended. In particular, all persons handling contaminated bodies must take appropriate precautions to minimize their dose, and the bodies should be treated in such a way as to minimize the spread of contamination. Many of these precautions are similar to those that would be taken under normal circumstances. 14.2 Guidelines for the Medical Examiner 14.2.1 Field Activities One of the duties of a medical examiner is to investigate the cause of death. This is especially important in the event of a criminal activity such as a terrorist attack, and a proper examination will require the medical examiner to perform this work in a radionuclidecontaminated area. Also the radiation-safety professionals (health physicists) who are present must understand the necessity for observing both radiological and nonradiological considerations when responding to many incidents involving radiation. For example, in an RDD attack, there is a need to preserve evidence of criminal 9

In addition, these documents address small numbers of decedents, not large numbers who might be expected following a terrorist attack or a major accident. A large-scale incident, especially a criminal act such as a terrorist attack, may place different demands on medical examiners and morticians; the actual radiation-safety actions taken will, of course, reflect the nature and severity of radiological risks as well as the demands of the particular incident. 10Levels of activity that may be considered relatively minor can produce exceptionally high levels of contamination. For example, a few kilobecquerels (~1 μCi) of activity is considered a relatively small amount of administered activity, considering that many radiopharmaceutical doses are on the order of many gigabecquerels (tens to hundreds of millicuries). But a few kilobecquerels (~1 μCi) will yield over two million disintegrations per minute, an amount that if not controlled can result in levels on skin (e.g., many times the 1,000 dpm considered a relatively-high level of skin contamination).

246 / 14. STAGE 9: CONTAMINATED DECEDENTS activity. Health physicists are not trained in such matters while medical examiners are not trained in radiation safety; both sets of skills are needed when recovering the remains of victims of a radiological attack. Medical examiners need not be radiation workers to enter a radionuclide-contaminated area provided they are accompanied by a health physicist or other radiation-safety worker. In addition, medical examiners should be aware of basic radiation-safety practices, including: • Stay-time in a radionuclide-contaminated area will be determined by the total dose to the most highly exposed person rather than the radiation dose rate to any individual at a given time. • Very-high radiation or contamination levels may delay or prevent entry into a radiological area. • The presence of very-high levels of radioactive contamination may pose an inhalation or ingestion concern due to resuspension, especially if it is necessary to work near ground level, where resuspended radionuclides may be most concentrated. • Elevated contamination levels may require wearing anticontamination clothing, such as shoe covers, protective gloves, coveralls, and respiratory protection (such as an N9511 mask, air purifying respirator, or forced air). • Exiting a contaminated area should be done according to the process outlined in Sections 5 and 6 of this Report. • Final determinations regarding stay-time, entry and exit procedures, anti-contamination clothing, and other radiationsafety precautions shall be the responsibility of the radiation worker, and persons from the office of the medical examiner shall follow the required radiation-safety precautions. • It may be necessary and appropriate to delay retrieval of decedents if high radiation or contamination levels prevent working safely in a radiological area. Radiation-safety workers accompanying medical examiners must remain aware they are providing coverage for work that may include elements of law enforcement, medical investigation, and that the work may have legal implications. Accordingly, it is important that 11By NIOSH classification, an N95 mask traps 95 % of particles with a diameter of 0.3 μm or larger. Any mask including N95 should be tested to ensure a good fit prior to wearing.

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radiation-safety personnel not take actions that are unapproved by the medical examiner on the scene; otherwise important evidence may be lost or destroyed, compromising the integrity of both medical and criminal investigations. In particular, radiation-safety (healthphysics) personnel must understand the following (Wood et al., 2007): • decedents’ bodies shall not be moved until authorized by the medical examiner; • decedents’ clothing and personal effects shall not be disturbed until authorized by the medical examiner; • decedents shall not be decontaminated or subjected to any contamination control measures until authorized by the medical examiner; • decedents’ bodies, all body parts, and body bags shall be labeled with the radiation symbol and/or a warning tag or label; • if possible, a field morgue should be used to minimize contamination of fixed facilities (such as a permanent morgue or funeral home); • if a field morgue is used, consider placing a refrigeration unit at least 10 m from the work area, to be used for all remains that are >1 mSv h–1 (100 mR h–1) on contact (i.e., at a distance of 1 cm from the surface); and • remains that are contaminated to levels of less than ~50 Bq (3,000 dpm) may be sent to a fixed facility. 14.2.2 Autopsy Decedents may be autopsied by the medical-examiner’s staff, particularly if the contamination was the result of criminal or negligent activities, or if this examination can help to shed light on the incident. Although the spread of contamination to internal organs poses no risk to the decedent at this point, identifying internal activity is an important part of the investigation as it can indicate whether the decedent was exposed to airborne activity or activity in a soluble chemical form. The spread of contamination from the skin (perhaps via contaminated surgical instruments) could lead to a misinterpretation of the physical and chemical characteristics of the contaminants, causing actions to be taken that later are seen to be unnecessary. It is possible that a decedent may contain radioactive shrapnel or other embedded radioactive material from an explosion. In such cases, it is possible that radiation levels may be sufficiently high so as to warrant precautions to avoid exceeding radiation dose limits or, in some cases, to avoid radiation injury.

248 / 14. STAGE 9: CONTAMINATED DECEDENTS The following recommendations for medical examiners should serve to reduce the risks of spreading contamination to personnel, the autopsy room, and to the internal organs; and to reduce the risk of excessive radiation dose to examining personnel. • The physician who pronounces the patient dead should attach a tag or note indicating the patient is radioactively contaminated. • Hospital and medical-examiner personnel working with contaminated cadavers should be trained in basic radiationsafety precautions. • Paperwork attached to the death certificate should indicate that the body is radioactively contaminated. If the decedent is known or suspected to have been struck with radioactive shrapnel, dose-rate measurements shall be performed prior to beginning an autopsy. Ideally, such measurements should have been performed, at the latest, shortly after admission; if so, the results should be noted on the chart. However, if such information is lacking, radiation-safety or medical-examiner staff must perform a radiation-level survey prior to conducting an autopsy. Examples of situations in which a person might contain embedded radioactive materials include (but are not limited to) proximity to the explosion of a radiological weapon, proximity to a steam explosion in a nuclear power plant, or status as a cancer therapy patient (with implanted radioactive sources) involved in an unrelated fatal incident. If possible, a trained radiation-safety professional should conduct radiation surveys or should interpret the results to provide appropriate guidelines prior to commencing an autopsy. If a trained radiation-safety professional is not available, trained personnel from nuclear medicine or radiation oncology may be able to perform this function. However, routine surveys may be performed by any properly trained radiation worker. These may include laboratory technicians and (in a research hospital) scientists, graduate and post-doctoral students, and research laboratory technicians. Radiation workers are limited to an annual radiation dose of 0.5 Sv (50 rem) to the skin or extremities (ICRP, 2007). When planning an autopsy (or multiple autopsies) it may be necessary to consider the measured radiation dose rate, the location of the shrapnel, the number of autopsies to be performed, and the number of trained persons in the medical-examiner’s office available to perform autopsies in order to determine the appropriate protective measures to be taken to prevent exceeding this dose limit. Examples of protective actions may include (but are not limited to):

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• removing the radioactive source and placing it into a shielded lead container (“pig”). However, this action carries with it some degree of risk and should be undertaken only by trained personnel and/or after suitable planning and rehearsal; • using multiple medical examiners and staff to quickly remove the organ(s) of interest to reduce the exposure time to any individual; • marking the location of the highest radiation dose rates (presumably in the areas nearest the shrapnel) so that less time is spent working in this area; • designating allowable “stay-times” for hands within the body cavity to avoid exceeding allowable doses to the extremities; • using extremity dosimeters (e.g., “ring badges”) to monitor radiation dose to the hands and fingers; • using self-reading whole-body dosimeters to monitor radiation dose to the person(s) performing the autopsy; • handling an organ removed during an autopsy as radioactive material if it contains radioactive fragments and disposing of it properly as radioactive waste when the autopsy is completed. The laboratory within which such organs are examined may require posting as a radioactive-materials storage area and/or a radiation area while the organ is present; • shielding properly any radioactive fragments removed from an organ or cadaver and transferring them to a competent authority at the earliest opportunity. Examples of such an authority include the hospital’s radiation-safety office, a local radioactive-materials licensee who is permitted to possess the radioactive material removed, and the appropriate regulatory authority. Alternately, the appropriate regulatory authority may permit the removed radioactive material to be retained onsite provided they are transferred for disposal promptly and are stored appropriately in the interim; • designating one person to remain “clean” by avoiding contact with the body or with equipment used for the autopsy if staffing permits. This person can be used to conduct radiological surveys, fill out paperwork, get supplies, and so forth; • covering the autopsy table with plastic and absorbent materials to reduce the chance of contamination; and • following these procedures and precautions as possible and appropriate if the decedent is known to be contaminated but does not contain radioactive shrapnel: - equipment such as pans, scales, trays, etc. that may come in contact with contaminated skin, organs, or body fluids

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may be covered with plastic and absorbent materials prior to use to reduce the chance of contamination. These coverings should be changed after each autopsy; when possible, the body should be put in a body bag or wrapped in plastic sheeting or blankets to reduce the spread of contamination; and if an autopsy is performed, all materials and equipment (including the gloves, surgical gowns, etc. worn by the autopsy staff) used in the autopsy should be considered radioactive materials and disposed of accordingly.

If possible, hospital or medical-examiner personnel should perform light decontamination prior to making their first incision. This will help to reduce contamination of surgical instruments and will lessen the risk of spreading contamination to internal organs. Decontamination efforts include wiping contaminated skin with a damp cloth or sponge, removing contaminated hair, flushing orifices with water, covering the decedent’s hands with plastic bags, and removing contaminated clothing. All clothing, contaminated hair, and decontamination materials (including liquids) should be disposed of as radioactive waste. All persons involved in the autopsy should make every reasonable effort to prevent spreading skin contamination into the body cavity or to internal organs. These efforts may include using different sets of surgical instruments for internal and external work, flushing incisions with water prior to penetrating the layers of skin and muscle, changing gloves following the initial incisions, etc. Upon completion of the autopsy, all instruments, drapes, scales, gloves, gowns, masks, and other equipment and protective clothing used must be considered to be radioactively contaminated and treated as such until surveyed and released for further use. This survey should be performed by a qualified radiation worker if possible. If scans indicate the absence of internal contamination, equipment that contacted only internal organs may be considered radiologically “clean” and may be reused. Following completion of the autopsy, the participating personnel should remove their outer garments and dispose of them as potentially-contaminated waste. They should then perform a whole-body contamination survey. If time does not permit a whole-body survey, they must, at a minimum, survey their hands, feet and face and decontaminate as necessary. Personnel participating in the autopsy may wish to consider submitting bioassay samples for analysis following completion of the autopsy.

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14.3 Guidelines for Mortuary Personnel The primary concern of mortuary personnel should be to minimize the spread of contamination to their facility, to personnel, and to the environment. Accordingly, the following guidelines emphasize contamination control measures. The presence of heavy contamination may call for closed casket funeral services, wrapping the body in plastic (or leaving it in a body bag), and other contamination control measures. Only in very rare cases where bodies are heavily contaminated with gamma-emitting radionuclides would lead-lined caskets be required. • Mortuary personnel working with radioactive cadavers should be trained in basic radiation-safety precautions. • If possible, radioactive sources, shrapnel, or other embedded radioactive materials should be removed from the body prior to transfer to the mortuary. If this is not possible, mortuary personnel should follow the precautions noted above. • The embalming table and facilities should be covered with plastic and disposable absorbent materials to the maximum extent possible. • Mortuary personnel should conduct mild decontamination, if possible, prior to moving the body to the mortuary. This may include wiping contaminated skin with a damp cloth or sponge, removing contaminated hair, flushing orifices with water, covering the decedent’s hands with plastic bags, and removing contaminated clothing. All clothing, contaminated hair, and decontamination materials (including liquids) should be disposed of as radioactive waste. • Plastic should be placed on floor beneath the embalming table to minimize the spread of contamination to the floor. • Wear appropriate protective clothing, to consist of double gloves, surgical mask, surgical gown (or “bunny suit” or the equivalent), shoe covers, face shield, and impermeable sleeves. • Protective clothing, the decedent’s clothing, fluids or sponges used to clean the body, dressings, trocars and tubing, and other materials that contact the body must be treated as radioactive waste. • Closed-cycle embalming should be used when possible to avoid the discharge of radioactive fluids. This may lead to the contamination of funeral home equipment, but will reduce the discharge of radioactive materials to the environment. • Body fluids and internal organs are not likely to be contaminated, with the possible exception of the lungs and related

252 / 14. STAGE 9: CONTAMINATED DECEDENTS fluids of patients who have inhaled large amounts of radioactivity. • A container should be dedicated to radioactive waste, and should be clearly marked as such with the radiation symbol or a sign. 14.4 Final Disposition of the Decedent • In general, cremation is not recommended. However, if cremation is decided upon, it may be advisable to decontaminate the body prior to cremation if contamination levels are high. Materials used for decontamination must be disposed of as radioactive waste [see ICRP Publication 94 (ICRP, 2004) and Publication 98 (ICRP, 2005b)] for international recommendations on cremation of bodies containing radionuclides). • The mortuary director must be advised that cremating contaminated patients may lead to contaminating the crematory. If the body is contaminated with long-lived radionuclides, this contamination may require treating the crematory materials (including the refractory brick inside) as radioactive material or as radioactive waste when the refractory brick is replaced. • Many forms of refractory brick utilize zirconium minerals because of their high melting temperature. However, such minerals are almost invariably associated with minor (and regulatory exempt) concentrations of uranium and uranium decay series nuclides. Because of this, it might be necessary to perform gamma spectroscopy or other isotope identification to determine the relative contributions of naturallyoccurring radioactive material versus radioactive material from contaminated decedents. • It may be necessary to delay cremations until an assessment of radiation dose to members of the public can be performed. For example, volatile radionuclides such as 137Cs or 131 I may be emitted with crematory exhaust gases, exposing people living or working in the area to radiation. The resulting radiation dose to members of the public is likely to be low, but regulations may require it be calculated and documented to confirm that members of the public were not exposed above regulatory limits. • Some factors will work to reduce radiation dose to members of the public. These include a high flow rate through the crematory exhaust, high crematory temperatures, low population in the vicinity of the funeral home, greater distance to

14.5 RELIGIOUS AND CULTURAL CONSIDERATIONS

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the nearest home, high average annual wind speed, and a typical wind direction that avoids nearby residents. Scattering of ashes into the environment should be delayed for 10 half-lives of the contaminating radionuclide, if possible, so that radioactive decay will reduce radiation dose rates. This may not be possible if the decedent was contaminated with long-lived radionuclides, in which case highactivity concentrations may preclude scattering ashes. Mortuary personnel working with contaminated bodies may wish to consider an appropriate bioassay (e.g., urine, feces and thyroid) following their work. A closed casket ceremony without viewing hours is advised if the decedent was heavily contaminated or if radioactive shrapnel is present and was not removed. No other radiological restrictions are suggested for the normal burial of contaminated decedents. 14.5 Religious and Cultural Considerations

Some religions require preparation of the body prior to burial, or may require burial within a given period of time. Other religions may require specific burial or disposition practices that may be contrary to good radiation-safety practices. When possible, these religious practices should be permitted because it is unlikely that a relatively short period of exposure to a radioactive or contaminated body will cause a harmful radiation dose or exceed dose limits. Decontamination of the body and removal of radioactive shrapnel facilitates these practices and should be performed when possible.

15. Contamination Control in Medical Facilities Objectives • minimize spread of radionuclide contamination from patient to staff and facilities; and • decontaminate personnel, equipment and facilities. 15.1 Introduction Very few hospital emergency departments are equipped to handle patients contaminated with radionuclides, even in those hospitals having radiation medicine departments. Major emphasis should be placed on controlling the contamination, by preventing radionuclides from being spread throughout the department and the rest of the hospital. Such a spread can happen very easily and quickly through inexperience and carelessness of the staff and the response personnel accompanying contaminated individuals. Generally, normal hospital sanitation practices will be adequate to control radionuclide contamination. However, radionuclides can present special problems. The following guidance is directed towards minimizing the spread of contamination from patients to staff and the facilities and to understanding proper procedures for decontamination of patients, personnel, equipment and facilities. Concern for contaminating equipment and facilities should not take precedence over treatment of contaminated patients. Equipment and facilities can be decontaminated. 15.2 Standard Precautions • Barrier protection should be used at all times to prevent skin and mucous membrane contamination by blood, body fluids containing visible blood, or other body fluids. • Gloves are to be worn when there is potential for hand or skin contact with blood, other potentially-infectious material, or items and surfaces contaminated with these materials. Double gloves should be considered, depending upon the situation 255

256 / 15. CONTAMINATION CONTROL IN MEDICAL FACILITIES • Face protection (face shield) is needed during procedures that are likely to generate droplets of blood or body fluid. • Protective body clothing (disposable clothing if possible) should be used when there is a potential for splashing of blood or body fluids. • Wash hands or other skin surfaces thoroughly and immediately if contaminated or if contamination is suspected. • Wash hands immediately after gloves are removed. • Avoid accidental injuries when working with sharp instruments or around sharp objects. • Place used needles, disposable syringes, scalpel blades, pipettes, and other sharp items in puncture-resistant containers. 15.3 Contamination Control Actions in the Emergency Department for Highly-Contaminated Patients • Wear personal protection equipment (PPE) and respiratory protection when treating patients: - surgical gloves - N95 mask or equivalent12 - shoe covers - coveralls Wearing respiratory protection is optional if exposure of medical personnel to airborne radionuclides is minor or of short duration. The highest priority should be maintaining effective communication between patient and medical personnel. • Lay down impermeable plastic floor covering if possible to establish a contamination control corridor directly from emergency department entrance to treatment rooms. • Move stretchers and gurneys along the contamination control corridor whenever possible. • Use dedicated rooms for all contaminated patients to minimize contaminating other parts of the hospital. • Leave controlled areas only at contamination control checkpoints. • Remove PPE and be surveyed by radiation protection personnel and be cleared to leave the controlled area. 12By NIOSH classification an N95 mask is a mask that traps 95 % of particles with a diameter of 0.3 μm or larger. Any mask, including an N95 mask should be tested to ensure a good fit prior to wearing.

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15.4 Working with Contaminated Patients • Treat life-threatening injuries first. • Obtain a nasal swab before decontaminating a patient (Section 9.3.1). This may not be possible in mass contamination situations. • Avoid spreading contamination into open wounds: - rinse with saline or deionized water. Clean skin (not wounds) with alcohol wipes if possible. • Wrap heavily contaminated patients in sheets or blankets. • Remove patient’s clothing if time permits. If this is not possible, the attending staff should dress in coveralls or surgical scrubs. • Use disposable equipment (e.g., blood pressure cuffs) when possible. • Assume that all equipment is radioactively contaminated if used on a contaminated patient: - decontaminate objects and equipment before using with other patients if possible; but - use without decontamination if necessary for lifesaving. 15.5 Hospital Emergency Department Contamination Controls A stylized diagram of an emergency department designed to receive contaminated patients is given in Figure 15.1. • Guidelines for controlling contamination: - Bring contaminated patients in through a point of access such as a fire exit other than the main entrance, where possible, to minimize contamination of the hospital. - Use textured plastic sheeting to cover the floors wherever possible. - Establish boundaries between inner contaminated, outer contaminated, and secured areas. These boundaries may be marked by tape applied to the floor. • Use a contamination control area (see Section 8 for decontamination procedures): - Decontaminate patients when their medical condition permits. - Perform decontamination in the treatment room which will likely be designated as being in the inner contaminated area whenever possible.

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Fig. 15.1. Stylized diagram of an emergency department and treatment-room complex to receive radioactively-contaminated patients. Dashed lines indicate radiological boundaries.

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Consider all items in the treatment room to be radioactively contaminated following admission of the first contaminated patient. Remove heavily contaminated clothing and cut hair as needed because significant amounts of contamination can adhere to a patient’s clothing and hair. Remove contaminated clothing in a cephalocaudal direction (i.e., away from the head). Eighty to 90 % of contamination is usually removed with the clothing (AFRRI, 2003; Bushberg et al., 2007; Koenig et al., 2005). Decontaminate patients and have them surveyed by a radiation-safety professional prior to being moved into a “clean” area unless their medical condition requires otherwise. Ensure that all staff take contamination control measures prior to exiting the contaminated area (Section 5.3).

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15.6 Contaminated Patients’ Rooms Figure 15.2 illustrates how treatment rooms can be set up for handling radioactively-contaminated patients. • Guidelines for controlling contamination: - use impermeable material (plastic) to cover floor in the outer contaminated area; - use a “sticky mat” or similar material for the step-off pad if available; and - mark the “hot” waste container with the radiation symbol and the words “radioactive waste” if possible. Line the container with a plastic bag to minimize contamination. • Use of contamination control area: - survey persons exiting a contaminated patient’s room (or have them surveyed) and decontaminate as necessary in accordance with Section 5.3; - dispose outer clothing (gloves, shoe covers, surgical gowns) as “hot” waste prior to leaving a patient’s room; and - do not reuse medical equipment that comes in contact with a contaminated patient (stethoscopes, blood pressure cuffs, etc.) unless it is first surveyed and found to be uncontaminated (except if needed immediately in lifesaving situations). 15.7 Patient Decontamination (Section 8) • Obtain a nasal swab prior to decontamination efforts if internal contamination is suspected (this may not be possible in mass contamination situations). • Remove patient’s clothing. • Rinse contaminated areas with saline solution or deionized water. • Patients should shower or bathe themselves unless seriously injured, using mild soap and cool to warm water or take a sponge bath, discarding sponges and washcloths as radioactive waste. (If patients are unable to shower or take a sponge bathe, medical personnel should assist them.) • Flush open wounds with saline solution or sterile water (Section 8.1.2). • Use standard sterile practices prior to administering injections, suturing, or other practices that puncture or break the skin.

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Fig. 15.2. Stylized map of treatment rooms set up for handling radioactively-contaminated patients.

15.8 Responsibilities of Radiation-Safety Personnel • Survey all patients prior to their entry into medical facilities. • Assist with patient decontamination when practicable. • Assist with establishing controlled areas for patient transport and treatment. • Survey controlled areas frequently to determine need for replacing or renewing coverings. • Establish and perform confirmation surveys of boundaries delineating controlled areas. • Survey medical and emergency personnel prior to exiting from controlled areas. • Perform bioassay measurements as necessary by collecting nasal swabs, urine, and fecal samples for analysis to determine possible uptakes of radionuclides by medical and emergency-response personnel. If intakes are suspected, in vivo and/or in vitro analyses may be recommended. • Perform bioassay measurements as necessary for patients thought to be internally contaminated with radionuclides by collecting nasal swabs, urine, and fecal samples for analysis. In vivo and/or in vitro analyses may be recommended. Identify the contaminating radionuclide(s). • Keep medical staff informed of contamination status of both the patients and the facility. 15.9 Hospital Decontamination Procedures for Protection of Personnel and Facilities Treatment of persons contaminated with radionuclides will result in some level of contamination of hospital emergency departments. Adherence to the following guidelines by radiation-safety,

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medical, and custodial personnel will help to minimize environmental and personal contamination and prevent radionuclides from being transported beyond established barriers. • Wear surgical scrub suits, caps and gowns, and rubber gloves (surgical, household or industrial, depending upon duties). • Train team leaders to recognize the rare instance when there may be a need for masks or respirators due to the presence of high levels of alpha or beta radionuclides. • Use rubber or plastic shoe covers if possible. Those performing decontamination with water should wear plastic or rubber laboratory aprons. • Turn off unfiltered air conditioning and forced-air heating systems to prevent radioactive dusts or aerosols from being carried into ducts or to other rooms. • Protect floors with disposable coverings (plastic or heavy paper) to reduce “tracking” of the material to other locations. The covering should be changed when significant contamination is present. It may be advantageous to run the covering three to four feet up the walls if practicable. • Remove shoe covers worn in contaminated areas when leaving the areas. Complete radiological surveys of persons leaving the areas should be conducted. • Place all contaminated clothing into plastic or paper bags carefully to reduce secondary contamination of the area. • Avoid splashing of irrigation fluids used in decontamination. • Move patients and other potentially-contaminated personnel to clean areas only after surveys show decontamination procedures have reduced the contamination to established control levels. • Survey and regulate all passage of persons and property between contaminated and clean areas using monitoring teams. • Pass supplies through monitoring stations from clean areas to contaminated areas. Passage out of contamination areas must not occur unless supplies are monitored and found clean. • Train all individuals on the decontamination team in radiological monitoring and in decontamination techniques. • Obtain fiberboard or steel drums with tight-fitting lids to contain and possibly store contaminated materials. Labels describing the contents should be affixed by radiation protection personnel so that proper disposal can be carried out without reopening the drums.

262 / 15. CONTAMINATION CONTROL IN MEDICAL FACILITIES • Personal dosimeters (pocket chambers or TLDs should be provided to all personnel working in the decontamination area. Personnel should be rotated out of the contaminated areas after receiving an effective dose of 50 mSv (5 rem), or less if possible [ICRP guidance for operations, including recovery and restoration after a radiation incident is an effective dose of 50 mSv (5 rem) (ICRP, 2005a)]. Restrict the entry of all nonessential personnel, family, visitors, administrative personnel, media, etc.

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Index blood samples 144, 150 fecal samples 144, 150 nasal swabs 144–146 tissue specimens 144, 151 urine samples 144, 146–148 Bioassay procedures, in vivo 151–153 chest (lung) counting 152 particular organs or tissues 152–153 whole-body counting 152 Bismuth 65–69, 182–184

Actinium 57, 65–69, 110, 154, 165, 182–186, 191–198, 233 absorbed doses 154 bioassay at 1 CDG 165 effective dose 154 inhalation dose 57 medical treatment 65–69, 182–186, 191–198, 233 Acute radiation syndromes 95, 99–100 Air-kerma rates 47–48 Americium 48, 53, 55, 57, 59, 61, 65–69, 110, 134, 143, 146–147, 149, 153–154, 157, 162, 165, 168, 182–186, 191–198, 233 absorbed doses 154 air-kerma rate 48 bioassay at 1 CDG 61, 165 deterministic effects, air concentrations 55 effective dose 154 electron constant 48 inhalation dose 57 medical treatment 65–69, 182–186, 191–198, 233 stochastic effects 53 Antimony 65–67, 182–184 Arsenic 65–67, 182–184

Cadmium 65–67, 182–184 Calcium 65–67, 182–184, 210, 212 Californium 48, 57, 65–69, 110, 154, 162, 165, 182–186, 191–198, 233 absorbed doses 154 air-kerma rate 48 bioassay at 1 CDG 165 effective dose 154 electron constant 48 inhalation dose 57 medical treatment 65–69, 182–186, 191–198, 233 Cancer statistics 15 Cancer surveillance 239–240 Carbon 65–69, 182–184 medical treatment 65–69, 182–184 CDG use, worked examples 171–175 60Co inhalation, adult 171 137 Cs exposure, child 172–174 137Cs inhalation, adult 171–172 137 Cs inhalation, population exposure 174–175 239 Pu inhalation, worker 174

Barium 65–67, 182–184, 210, 212 Berkelium 65–67, 182–184 Bioassay predictions, intake of 1 CDG 163–166 chest retention 163–166 nasal swab 163–166 urinary excretion 163–166 whole-body retention 163–166 Bioassay procedures, in vitro 143–151

277

278 / INDEX Cerium 51, 53, 55, 57, 65–69, 110, 143, 154, 164, 182–186, 191–198, 233 absorbed doses 154 bioassay at 1 CDG 164 deterministic effects 51 deterministic effects, air concentrations 55 effective dose 154 inhalation dose 57 medical treatment 65–69, 182–186, 191–198, 233 stochastic effects 53 Cesium 45, 47, 51–53, 55, 57, 61, 65–69, 110, 129, 143, 146, 154, 162, 164, 168, 171–175, 178, 181–186, 201, 203–209, 234, 252 absorbed doses 154 air-kerma rate 47 bioassay at 1 CDG 61, 164 deterministic effects 51 deterministic effects, air concentrations 55 effective dose 154 electron constant 47 inhalation dose 57 medical treatment 65–69, 182–186, 201, 203–209 stochastic effects 53 Check list 119, 136–137 medical information 136–137 supplies, patient decontamination 119 Chest (lung) counting 152, 163–166 actinides, low-energy photons 152 data interpretation factors 152 detectors 152 locations 152 low backgrounds 152 related to 1 CDG 163–166 Chromium 65–67, 182–184 Clinical Decision Guide (CDG) (see Stage 6)

Cobalt 32, 47, 53, 57, 61, 62, 65–69, 110, 154, 162–163, 171–172, 182–186, 191, 199, 201–202, 210, 212–213 absorbed doses 154 air-kerma rate 47 bioassay at 1 CDG 61, 163 effective dose 154 electron constant 47 inhalation dose 57 medical treatment 65–69, 182–186, 212–213 stochastic effects 53 Contaminated decedents (see Stage 9) Contamination control areas 3, 76–78, 89–91, 257–261 inner contaminated area 76–78 onsite 3, 76–78 outer contaminated area 76–78 secured area 76–78 Contamination control, basic principles 88–92 controlled contamination areas 89–91 equipment 92 exposed persons 88–89 influencing factors 88 medical and emergency responders 89 transportation 91–92 Contamination control, medical facilities 113–120, 255–262 contaminated patients 257 contamination control complex 257–258 decontamination, personnel, facilities 260–262 highly contaminated patients 256 objectives 255 patient decontamination 113–120, 259 patients’ rooms 259 personal dosimeters 262 proper waste disposal 261 radiation-safety personnel 260

INDEX

standard precautions 255–256 Contamination, external, individuals 101, 107–121 assessment 107–112 decontamination 113–121 screening 101 Copper 65–67, 182–184 Curium 57, 65–69, 110, 154, 165, 181–186, 191–198, 233 absorbed doses 154 bioassay at 1 CDG 165 effective dose 154 inhalation dose 57 medical treatment 65–69, 182–186, 191–198, 233 Decedents, radionuclide contaminated 244–253 Decision tree, radionuclide contaminated persons 39 Decontamination, external, individuals (see Stage 3) Decontamination facilities 41, 80, 90–96, 113, 118, 257–259 Decontamination supplies 119 Decontamination, wound 132–133 Decorporation therapy, by drug 187–211 BAL, dimercaprol 189–190 Ca-DTPA and Zn-DTPA 191–198 DFOA, deferoxamine 187–189 DNSA, succimer 209–210 EDTA 198–200 penicillamine 199, 201–202 Prussian blue 201, 203–209 Decorporation therapy, overview 65–69, 179–187 dose schedules, drug or treatment 65–69, 182–184 FDA-approved drugs 181 FDA guidance 180–187 therapy recommendations, by element 65–69, 182–184 Dermal injury 5, 107, 111, 113–114, 123–124, 131–132 Deterministic effects, air concentrations 55

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Deterministic health effects 1–2, 6–7, 14, 18, 27, 29, 46, 49, 51, 54–56, 60, 114–115, 142–143, 159–161, 228, 258 Dose-rate measurements, significance 36 Dosimeters, personal (see Personal dosimeters) DTPA treatment 191–198 Ca-DTPA, Zn-DTPA 192–195 clinical experience 197–198 contaminated wounds 192 FDA indication 191 general information 195–197 summary of treatment recommendations 192–195 treatments considered by others 191 Electron constants 47–48 Emergency medical management, hospital 128–132 contaminated, life-threatening injuries 128 contaminated, lightly injured 128–129 contaminated, uninjured 129 radiation burns 131–132 radiation injury 129–131 Europium 57, 65–69, 154, 164, 182–186, 191–198 absorbed doses 154 bioassay at 1 CDG 164 effective dose 154 inhalation dose 57 medical treatment 65–69, 182–186, 191–198 Evaluation and emergency care, hospital (see Stage 4) Excretion sampling (see Bioassay procedures, in vitro) External contamination assessment (onsite triage) (see Stage 2) External decontamination, individuals (onsite) (see Stage 3)

280 / INDEX External radiation exposures 42–49 criticality accident 42, 44–45 sealed sources 45–48 skin contamination 46–49 whole-body exposure 42–43, 45–46 Fecal bioassay samples 148–150 collection 150 identification 150 normalization, sampling interval 149 sampling protocol 148 size 150 total voidings 150 First responders, guidance 74–79 control areas 76–78 first at scene 75 immediate protection goals 75–76 protection, first responders 78–79 First responders, radiation safety 72–79 general instructions 73–74 major objectives 72 teamwork needed 72 Fission products (mixed) 65–67, 182–184 Fluorine 65–67, 182–184 Follow-up medical care (see Stage 8) Gallium 65–67, 182–184 Gold 65–67, 182–184 Hydrogen (see Tritium) Incident details 140–143 location and time 140 physical and chemical form 142–143 radionuclide identification 142–143 route of exposure 141–142 Incident response 33–37

guidance for professionals 35–37 roles and responsibilities 34–35 scale of incident 33–34 Indium 65–67, 182–184 Information sources 12–14, 25–26 contact information 25–26 recent publications 12–14 Initial treatment decisions, hospital 132–136 Clinical Decision Guides 136 drug decorporation therapy 136 emetics 135 medical management algorithms 136 purgatives 135 radionuclide-contaminated wound 132–133 radionuclide ingestion 134–135 radionuclide inhalation 133–134 Intake and dose assessments 138–153 Internal contamination assessment, hospital (see Stage 5) Intervention levels, skin contamination 49–51 Iodine 5, 6, 38, 40–41, 47, 49, 51–53, 55–57, 60–61, 63–69, 104–105, 110, 134, 143, 153–154, 159–161, 165, 167, 179, 181–186, 212, 214–221, 252 absorbed doses 154 air-kerma rate 47 bioassay at 1 CDG 64, 167 deterministic effects 51 deterministic effects, air concentrations 55 effective dose 154 electron constant 47 inhalation dose 57 medical treatment 65–69, 182–186, 212, 214–221 stochastic effects 53 Iridium 47, 57, 61, 65–69, 110, 155, 161–162, 164, 182–186, 191–198 absorbed doses 155

INDEX

air-kerma rate 47 bioassay at 1 CDG 61, 164 effective dose 155 electron constant 47 inhalation dose 57 medical treatment 65–69, 182–186, 191–198 Iron 65–67, 182–184 KI treatment [see Potassium iodide (KI) treatment] Lanthanum 65–67, 182–184 Lead 65–67, 182–184 Lung lavage 233–236 benefit-to-risk assessment 235–236 criteria for use 233–234 human case 235 procedure 234 results in laboratory animals 234–235 Magnesium 65–67, 182–184 Management stages 4–7, 16–21, 38–42, 93–253 Stage 1: Medical assessment 4, 40, 93–106 Stage 2: External contamination assessment 5, 40, 107–112 Stage 3: External decontamination 5, 41, 113–121 Stage 4: Patient evaluation and emergency care 5, 41, 123–137 Stage 5: Internal contamination 5, 41, 138–157 Stage 6: Clinical decision guidance 6, 41, 158–175 Stage 7: Medical management 6, 41, 176–236 Stage 8: Follow-up medical care 7, 42, 237–243 Stage 9: Contaminated decedents 7, 42, 244–253 Manganese 65–67, 182–184

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Medical assessment (onsite triage) (see Stage 1) Medical examiner, autopsy 247–250 contamination control, facility 247–250 contamination control, personal 247–250 contamination surveys 247–250 embedded radioactive material 247–249 personal dosimeter 249 radiation-safety measures 247–250 radiation survey 247–249 shielding 249 stay-time control 248–249 Medial examiner, field activities 245–247 contamination control 245–247 contamination surveys 245–247 law-enforcement issues 246–247 personal protection equipment 246 radiation-safety assistance 245–247 radiation surveys 246–247 stay-time in contaminated area 246 Medical history 125, 239 Medical management (see Stage 7) Mercury 65–67, 182–184 MIPS (see Multiple idiopathic physical symptoms) Molybdenum 65–67, 182–184 Mortuary personnel guidelines 251–252 closed-cycle embalming 251–252 contamination control 251–252 embedded radioactive material 251–252 protective clotting 251–252 radioactive waste 251–252 Multiple idiopathic physical symptoms (MIPS) 241–242 mortality increases 241–242

282 / INDEX nonspecific somatic complaints 241 uncertain durations 241–242 variable outcomes 241–242 N95 mask 83, 87, 246, 249, 256 Nasal swabs 144–146, 163–166 analysis 145 collection 145 interpretation 144–146 related to 1 CDG 165–166 Neptunium 65–67, 182–184 Neutron dose, 24Na activation 42, 44–45 Nickel 65–67, 182–184 Niobium 65–67, 182–184 Notification of incident 33 Palladium 65–69, 110, 155, 163, 182–186, 199, 201 absorbed doses 155 bioassay at 1 CDG 163 effective dose 155 medical treatment 65–69, 182–186, 199, 201 Personal dosimeters 1, 4, 34–35, 79, 86–87, 96–97, 118–119, 249, 259 first responders 79, 86–87, 96–97 incident site 34–35 medical examiner 249 medical facility 259 onsite decontamination activities 118–119 Personal protection equipment (PPE) 63, 73–74, 78–79, 83, 86–91, 120, 256 appropriate types 86–87 dressing order 86 examples 83 inspections 83 need for 73–74 removal 86, 88 use by emergency responders 78–79

Phosphorus 58, 65–69, 110, 155, 161, 163, 169, 182–186, 221–223 absorbed doses 155 bioassay at 1 CDG 163 effective dose 155 inhalation dose 58 medical treatment 65–69, 182–186, 221–223 Plutonium 5, 30, 47, 51–53, 55, 58–59, 61–62, 65–69, 104–105, 110, 116, 133–134, 136, 143, 145–147, 149, 153–155, 157, 162, 165–166, 168, 174, 179–181, 182–186, 187, 191–198, 196–197, 233, 235–236, 239 absorbed doses 155 air-kerma rate 47 bioassay at 1 CDG 61, 165 deterministic effects 51 deterministic effects, air concentrations 55 effective dose 155 electron constant 47 inhalation dose 58 medical treatment 65–69, 182–186, 191–198, 233 stochastic dose 53 Polonium 30, 47, 51–53, 55, 58, 62, 65–69, 129, 143, 155, 164, 166, 169, 182–186, 189–190, 199–202, 210 absorbed doses 155 air-kerma rate 47 bioassay at 1 CDG 164 deterministic effects 51 deterministic effects, air concentrations 55 effective dose 155 electron constant 47 inhalation dose 58 medical treatment 65–69, 182–186, 189–190, 199–201 stochastic effects 53 Potassium 65–67, 182–184 Potassium iodide (KI) treatment 212, 214–221 additional information 215–221

INDEX

alternate therapy modes 214 possible side effects 220 recommended doses 214 treatment recommendations 212, 215 Preventive medicine approaches 239–240 age-dependent tests 239 cancer surveillance 239 follow-up bioassays 239–240 physician judgment 239 screening tools 239 Promethium 65–67, 182–184 Prussian blue treatment 201–209 clinical experience 204–209 efficacy 204 FDA approval 201 FDA indication 201 modes of treatment 201, 203–204 precautions 204 studies in laboratory animals 206, 207 Psychological distress 241–242 anger 241 difficulty concentrating 241 difficulty sleeping 241 disbelief 241 fear 242 multiple idiopathic physical symptoms (MIPS) 241–242 sadness 241 Psychosocial issues 240–243 acute phase 240 behavioral changes 241 distress 241 long-term follow-up 240 multiple idiopathic physical symptoms (MIPS) 241–243 risk of psychiatric illness 241 Psychosocial issues, affected groups 240–243 children 242 clinicians 242–243 exposed persons 240–243 first receivers 242 first responders 242 parents 242 Public health 8, 13, 35, 73, 239

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Radiation and contamination surveys 80–85 area survey, beta/gamma 82 contaminated surfaces, alpha 82 external radiation sources 80–85 instruments, survey types 84–85 radioactive contamination 80–85 survey guidance, general 81–82 Radiation control areas 3, 21, 76–78, 89–91, 257–261 examples of use 89–91 inner contaminated area 76–78 outer contaminated area 76–78 secured area 76–78 Radiation detectors and survey instruments 31, 35, 40, 42, 44–45, 81–82, 85, 103, 109, 152, 196 appropriate instruments 85 Radiation dose-limit recommendations 36, 63, 70–71, 78–79 FEMA, emergency 71 ICRP 70 lifesaving 36, 63, 71, 78–79 NCRP 70 Radioactive waste 88–91, 118–119, 249–252, 259 autopsy 249–250 incident site 88–91 medical facility 259 mortuary 251–252 onsite decontamination 118–119 Radiological facts, basic 30–33 Radium 30, 47, 52–53, 58, 61, 65–69, 110, 155, 164, 182–186, 221, 224–227 absorbed doses 155 air-kerma rate 47 bioassay at 1 CDG 61, 164 effective dose 155 electron constant 47 inhalation dose 58 medical treatment 65–69, 182–186, 221, 224–227 stochastic effects 53

284 / INDEX Rhenium 155, 164 absorbed doses 155 bioassay dose at 1 CDG 164 effective dose 155 Rubidium 65–67, 182–184 Ruthenium 53, 57, 58, 65–69, 110, 155, 163, 182–186, 191–198 absorbed doses 155 bioassay at 1 CDG 163 effective dose 155 inhalation dose 58 medical treatment 65–69, 182–186, 191–198 stochastic effects 53 Samarium 58, 156, 161, 164 absorbed doses 156 bioassay at 1 CDG 164 effective dose 156 inhalation dose 58 Scandium 65–67, 182–184 Screening, contamination 101–103, 111 SI and previous system units 23 Silver 65–67, 182–184 Skin burns 17, 41, 88, 101, 105, 111, 123–124, 127, 131–132 emergency treatment 123–124 medical management 131–132 patient evaluation 17, 41, 124, 127 treatment guidance 111 Sodium 65–67, 182–184 Stage 1: Medical assessment (onsite triage) 4, 17–19, 39–40, 93–106 acute radiation syndromes 96, 99–100 documentation needs 104 excreta collections 103 identifying exposed individuals 96 initial actions 94 intakes by ingestion or absorption 103 intakes by inhalation 102–103 intakes through wounds 103

life-threatening problems 94–96 objectives 93 onsite treatment, internal contamination 104 processing priorities, exposed persons 104–106 radiation exposure assessments 96–98 screening for external contamination 101 screening for internal contamination 101–103 Stage 2: External contamination assessment (onsite triage) 5, 17–18, 20, 39–41, 107–112 assessment procedures 107–109 intervention levels, skin 111–112 objectives 107 skin dose assessment 109–110 treatment guidance 111 Stage 3: External decontamination, individuals (onsite) 5, 17–18, 20, 39, 41, 113–121 decontamination of persons 113–114 documentation needs 121 facilities 118, 120 goals 114–116 guidance to responders 117–118 objectives 113 post-decontamination procedures 120–121 procedures 115–117 saving contaminated materials 120 supplies 118, 119 Stage 4: Evaluation and emergency care, hospital 5, 17–18, 20, 39, 41, 123–137 emergency medical management 128–132 external radiation 124–125, 129–132, 136–137 general instructions 127–128 information checklist 136–137 initial treatment 132–136

INDEX

internal contamination 124–125, 127–129, 132–137 objectives 123 psychosocial and behavioral aspects 125–126 Stage 5: Internal contamination assessment, hospital 5–6, 17–18, 20, 39, 41, 138–157 bioassay procedures, in vitro 143–151 bioassay procedures, in vivo 151–153 dose assessment procedures 138, 139 incident details 140 intake and dose assessments 153–157 objectives 138 Stage 6: Clinical Decision Guide (CDG) 6, 18, 20, 39, 41, 55–56, 60–64, 136, 142–143, 146, 149, 158–175, 179–180 bioassay predictions, intake of 1 CDG 162–167 CDG definitions for 131I 159–162, 167, 169 CDG definitions for nonradioiodine radionuclides 158–166 CDG use, worked examples 171–175 factors influencing CDG use 161–162 instrument-related considerations 169 normalization of urine values 149, 169–171 objectives 158 particle-size sensitivity 162, 168 radioiodine, FDA guidance 160–161 Stage 7: Medical management 6–7, 17–18, 20, 39, 41, 176–253 decorporation therapy, by drug 187–210 decorporation therapy, overview 179–187

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lung lavage 233–236 medical treatments, by radionuclide 210–253 objectives 176 therapeutic concepts 177–179 Stage 8: Follow-up medical care 7, 17–18, 20, 39, 42, 237–243 late-occurring health effects 238–239 objectives 237 preventive medicine approaches 239–240 psychosocial issues 240–243 Stage 9: Contaminated decedents 7, 17–18, 21, 39, 42, 244–253 decedent final disposition 252–253 medical examiner guidelines 245–250 mortuary personnel guidelines 251–252 objectives 244 Standard precautions 255–256 avoiding accidental injuries 256 barrier protection 255 face protection 256 gloves 255 hand washing 256 proper disposal, sharp items 256 protective body clothing 256 Stochastic health effects 1–2, 6–7, 14, 29, 52, 56, 60, 115, 142, 159–161, 237 Strontium 47, 51, 53, 55, 58, 61, 65–69, 110, 134, 143, 146, 156, 162–163, 168, 182–186, 221, 224–227 absorbed doses 156 air-kerma rate 47 bioassay at 1 CDG 61, 163 deterministic effects 51 deterministic effects, air concentrations 55 effective dose 156 electron constant 47 inhalation dose 58

286 / INDEX medical treatment 65–69, 182–186, 221, 224–227 stochastic effects 53 Sulfur 65–67, 182–184 Surface contamination readings, significance 37 Target audiences 15 Technetium 65–69, 156, 179, 182–186, 191, 201, 203–209 absorbed doses 156 effective dose 156 medical treatment 65–69, 182–186, 201, 203–209 Terminology, abbreviated list 26–29 Thallium 65–67, 182–184 Thorium 5, 52–53, 59, 65–69, 156, 165, 182–186, 191–198, 231, 233 absorbed doses 156 bioassay at 1 CDG 165 effective dose 156 inhalation dose 5 medical treatment 65–69, 182–186, 191–198, 233 stochastic effects 53 Training 2, 10, 12, 15, 22, 73, 144, 242 Triage, medical-radiological 17, 19, 20, 33, 40, 73, 78, 80, 93–94, 101, 107, 118, 124, 126 Tritium 53, 57, 65–69, 149, 154, 157, 163, 165–166, 169, 179, 182–186, 184, 228–229 absorbed doses 154 bioassay at 1 CDG 163 effective dose 154 inhalation dose 57 medical treatment 65–69, 182–186, 228–229 stochastic effects 53 Uranium 5, 47, 52–53, 59, 61–62, 65–69, 156–157, 165–166, 179, 182–186, 191, 193, 195–197, 229–233, 252

absorbed doses 156 air-kerma rate 47 bioassay at 1 CDG 61, 165 effective dose 156 electron constant 47 inhalation dose 5 medical treatment 65–69, 182–186, 229–233 Urine bioassay samples 146–149, 163–166 collection 149 identification 149 normalization, sampling interval 149 related to 1 CDG 163–166 sampling protocols 147–148 size 149 U.S. Strategic National Stockpile 197 Whole-body counting 152, 163–166 configurations 152 detectors 152 locations 152 low backgrounds 152 related to 1 CDG 163–166 Wound decontamination 132–133 (also see Stage 3 and Stage 4) Yttrium 5, 47, 59, 65–69, 110, 156, 163, 182–186, 191–198, 233 absorbed doses 156 air-kerma rate 47 bioassay at 1 CDG 163 effective dose 156 electron constant 47 inhalation dose 5 medical treatment 65–69, 182–186, 191–198, 233 Zinc 65–67, 182–184 Zirconium 65–67, 182–184