Population Monitoring and Radionuclide Decorporation Following a Radiological Or Nuclear Incident

166

NCRP REPORT No. 166

POPULATION MONITORING AND RADIONUCLIDE DECORPORATION FOLLOWING A RADIOLOGICAL OR NUCLEAR INCIDENT

POPULATION MONITORING AND RADIONUCLIDE DECORPORATION FOLLOWING A RADIOLOGICAL OR NUCLEAR INCIDENT

National Council on Radiation Protection and Measurements

NCRP REPORT No. 166

Population Monitoring and Radionuclide Decorporation Following a Radiological or Nuclear Incident

Recommendations of the NATIONAL COUNCIL ON RADIATION PROTECTION AND MEASUREMENTS

April 6, 2010

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 publications. 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 Population monitoring and radionuclide decorporation following a radiological or nuclear incident / recommendations of the National Council on Radiation Protection and Measurements. p. ; cm. -- (NCRP report ; no. 166) "April 6, 2010." Includes bibliographical references and index. ISBN 978-0-9823843-7-4 (alk. paper) 1. Radiation--Physiological effect. 2. Radiation--Safety measures. I. National Council on Radiation Protection and Measurements. II. Series: NCRP report ; no. 166. [DNLM: 1. Radiation Effects--Practice Guideline. 2. Decontamination--methods-Practice Guideline. 3. Disaster Planning--Practice Guideline. 4. Radiation Monitoring--Practice Guideline. 5. Radioactive Hazard Release--Practice Guideline. 6. Triage--Practice Guideline. WN 600] RC93.P67 2011 616.9'897--dc23 2011014172

Copyright © National Council on Radiation Protection and Measurements 2011 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. [For detailed information on the availability of NCRP publications see page 273.]

Preface In the years since the September 11, 2001 terrorism incident, the National Council on Radiation Protection and Measurements (NCRP) has been active in preparing publications that provide guidance in preventing, preparing for, and responding to possible acts of radiological or nuclear terrorism. Major publications on these subjects include Report No. 138, Management of Terrorist Events Involving Radioactive Material and Commentary No. 19, Key Elements of Preparing Emergency Responders for Nuclear and Radiological Terrorism. In addition, NCRP has published three commentaries (Commentary No. 16, Screening of Humans for Security Purposes Using Ionizing Radiation Scanning Systems; No. 17, Pulsed Fast Neutron Analysis System Used in Security Surveillance; and No. 20, Radiation Protection and Measurement Issues Related to Cargo Scanning with Accelerator-Produced High-Energy X Rays) that discuss measurement and health protection aspects of using radiation-based systems for the detection and interdiction of radiological or nuclear materials and terrorism threats. NCRP has also been preparing new reports related to the treatment and long-term medical management of people affected by deliberate or accidental releases of radiological or nuclear materials including Report No. 161, Management of Persons Contaminated with Radionuclides. The present Report provides guidance for emergency responders and medical centers for the development of radiological response plans that include the efficient screening of a population for internally-deposited radionuclides, decontamination procedures, and treatment by decorporation therapy. While this Report is intended to focus on screening a population for internal contamination, screening is only one aspect of monitoring a population in the aftermath of a radiological or nuclear incident. Thus, this Report also broadly discusses external monitoring and decontamination of the affected population. The social and psychological impacts of a radiological or nuclear incident, and the long-term medical monitoring of the exposed population, are also addressed. A focus is placed on rapid methods for determining the contaminating radionuclides and the screening, decontamination and prompt medical management of contaminated persons.

iii

iv / PREFACE This Report was prepared by NCRP Scientific Committee 4-2 on Population Monitoring and Decontamination Following a Radiological or Nuclear Incident. Members of the Committee were: Richard J. Vetter, Chairman Mayo Clinic Rochester, Minnesota Members Steven M. Becker University of Alabama at Birmingham Birmingham, Alabama

Steven H. King Milton S. Hershey Medical Center Pennsylvania State University Hershey, Pennsylvania

Eugene H. Carbaugh Pacific Northwest National Laboratory Richland, Washington

Adela Salame-Alfie New York State Department of Health Troy, New York

James R. Cassata U.S. Navy Bethesda, Maryland

Lin-Shen Casper Sun U.S. Nuclear Regulatory Commission Rockville, Maryland

Scott Davis University of Washington Seattle, Washington

Katherine Uraneck New York City Department of Health and Mental Hygiene New York, New York

Fun H. Fong, Jr. Smyrna, Georgia

George J. Vargo MJW Corporation, Inc. Avondale, Pennsylvania

P. Andrew Karam New York City Department of Health and Mental Hygiene New York, New York NCRP Secretariat Bruce B. Boecker, Staff Consultant Cindy L. O’Brien, Managing Editor David A. Schauer, Executive Director

The Council expresses its appreciation to the Committee members for the time and effort devoted to the preparation of this Report. NCRP also expresses appreciation to the Centers for Disease Control and Prevention and the U.S. Navy for providing funding for preparation of the Report. Thomas S. Tenforde President

Contents Preface . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii 1. Executive Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1 2. Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .7 2.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .7 2.2 Purpose of this Report . . . . . . . . . . . . . . . . . . . . . . . . . . . .13 2.3 Target Audiences of this Report . . . . . . . . . . . . . . . . . . . .14 2.4 Scope of this Report . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14 3. Background Information . . . . . . . . . . . . . . . . . . . . . . . . . . . . .17 3.1 Internal Deposition of Radionuclides . . . . . . . . . . . . . . . .17 3.1.1 Inhalation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .18 3.1.2 Ingestion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .20 3.1.3 Absorption from Skin Contamination . . . . . . . . .21 3.1.4 Absorption Through Wounds . . . . . . . . . . . . . . . .21 3.2 External Contamination . . . . . . . . . . . . . . . . . . . . . . . . . .22 3.3 Effects of Weather . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .22 3.4 Complications Due to the Presence of Multiple Agents or Serious Injuries . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .23 3.5 Radiological Triage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .26 3.6 Proximity to the Incident . . . . . . . . . . . . . . . . . . . . . . . . . .26 3.7 Previous Experience with Internal Contamination . . . . .27 3.8 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .28 4. Settings in Which Persons May Become Contaminated with Radioactive Material . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .29 4.1 Radiological Dispersal Device . . . . . . . . . . . . . . . . . . . . . .29 4.1.1 Incident Characteristics . . . . . . . . . . . . . . . . . . . .30 4.1.2 Nature of Contamination . . . . . . . . . . . . . . . . . . .30 4.2 Aerosol Dispersal into a Public Area . . . . . . . . . . . . . . . .31 4.2.1 Incident Characteristics . . . . . . . . . . . . . . . . . . . .31 4.2.2 Nature of Contamination . . . . . . . . . . . . . . . . . . .31 4.3 Contamination of Food or Water Supplies . . . . . . . . . . . .32 4.3.1 Incident Characteristics . . . . . . . . . . . . . . . . . . . .32 4.3.2 Nature of Contamination . . . . . . . . . . . . . . . . . . .33 4.4 Improvised Nuclear Device . . . . . . . . . . . . . . . . . . . . . . . .33 4.4.1 Incident Characteristics . . . . . . . . . . . . . . . . . . . .33 4.4.2 Nature of Contamination . . . . . . . . . . . . . . . . . . .35

v

vi / CONTENTS 4.5

Nuclear Reactor Incident . . . . . . . . . . . . . . . . . . . . . . . . . 4.5.1 Incident Characteristics . . . . . . . . . . . . . . . . . . . . 4.5.2 Nature of Contamination . . . . . . . . . . . . . . . . . . . Large-Scale Fires and Incidents . . . . . . . . . . . . . . . . . . . 4.6.1 Incident Characteristics . . . . . . . . . . . . . . . . . . . . 4.6.2 Nature of Contamination . . . . . . . . . . . . . . . . . . . Sealed Radioactive Source Incidents . . . . . . . . . . . . . . . . 4.7.1 Incident Characteristics . . . . . . . . . . . . . . . . . . . . 4.7.2 Nature of Contamination . . . . . . . . . . . . . . . . . . . Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

35 36 37 37 37 38 38 39 39 40

5. Coordination with the Incident Command System . . . . . 5.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2 Incident Command System . . . . . . . . . . . . . . . . . . . . . . . 5.3 Hospital Incident Command System . . . . . . . . . . . . . . . . 5.4 Coordination Between Incident Command System and Hospital Incident Command System . . . . . . . . . . . . . . . . 5.5 Communicating Information from the Scene to the Hospitals and from the Hospitals to the Scene . . . . . . . .

41 41 43 46

6. Radiological Triage and Screening Guidance . . . . . . . . . . 6.1 General Guidance for Emergency Responders . . . . . . . . 6.1.1 Selecting an Appropriate Radiation Survey Instrument . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.1.2 Presurvey Radiation Survey Instrument Checks 6.1.3 Surveying for Radioactive Contamination . . . . . 6.2 Radiological Triage and Screening Procedures . . . . . . . . 6.3 Initial Screening at Scene . . . . . . . . . . . . . . . . . . . . . . . . 6.4 Initial Screening at Hospital . . . . . . . . . . . . . . . . . . . . . . 6.5 Mass Screening Following the Emergency Phase . . . . . . 6.6 Biodosimetry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

53 53

4.6

4.7

4.8

48 48

54 55 56 56 57 60 61 61

7. Clinical Decision Guide: Concept and Use . . . . . . . . . . . . . 65 7.1 Clinical Decision Guide Concept . . . . . . . . . . . . . . . . . . . 65 7.2 Clinical Use of the Clinical Decision Guide . . . . . . . . . . . 68 7.2.1 Decision-Making Process . . . . . . . . . . . . . . . . . . . 68 7.2.2 Use of the CDG Tables . . . . . . . . . . . . . . . . . . . . . 69 7.2.3 Use of a Single-Void Urine Sample Collected During the First 24 h . . . . . . . . . . . . . . . . . . . . . . 70 7.2.4 Using the CDG with an Intake of Multiple Radionuclides . . . . . . . . . . . . . . . . . . . . . . . . . . . . 70 7.2.5 Determining an Intake for Times More than 24 h in the Past . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 71 7.2.6 Special Considerations for Uranium CDGs . . . . 73 7.3 Americium-241: Clinical Decision Guide Fact Sheet . . . 78

CONTENTS

7.4 7.5 7.6 7.7 7.8 7.9 7.10 7.11 7.12

7.13

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Cesium-137: Clinical Decision Guide Fact Sheet . . . . . . .79 Cobalt-60: Clinical Decision Guide Fact Sheet . . . . . . . . .80 Iodine-131: Clinical Decision Guide Fact Sheet . . . . . . . .82 Iridium-192: Clinical Decision Guide Fact Sheet . . . . . . .84 Plutonium-238: Clinical Decision Guide Fact Sheet . . . .85 Plutonium-239: Clinical Decision Guide Fact Sheet . . . .86 Radium-226: Clinical Decision Guide Fact Sheet . . . . . .88 Strontium-90: Clinical Decision Guide Fact Sheet . . . . .89 Uranium: Clinical Decision Guide and Nephrotoxicity Fact Sheet . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .90 7.12.1 Treatment Based on Radiological Properties of Uranium (CDG) . . . . . . . . . . . . . . . . . . . . . . . . . . .90 7.12.2 Treatment Based on Nephrotoxic Properties of Uranium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .91 Clinical Decision Guide Technical Details . . . . . . . . . . . .93 7.13.1 Noniodine Radionuclides . . . . . . . . . . . . . . . . . . . .93 7.13.2 Iodine Radionuclides . . . . . . . . . . . . . . . . . . . . . . .94 7.13.3 Uranium Mass and Activity Calculations . . . . . .94

8. Rapid Determination of Internal Contamination . . . . . . .96 8.1 Rapid Identification of Radionuclide(s) Involved . . . . . . .97 8.2 Screening for External Contamination . . . . . . . . . . . . . . .98 8.3 Direct (in vivo) Screening for Internal Contamination . .99 8.3.1 Detection of Internal Contamination by Direct Measurement . . . . . . . . . . . . . . . . . . . . . . . . . . . . .99 8.3.1.1 GM Survey Meter . . . . . . . . . . . . . . . . .101 8.3.1.2 Whole-Body and Lung Counters . . . . .101 8.3.2 Hospital Equipment for the Detection and Quantitation of Radionuclides . . . . . . . . . . . . . .103 8.3.2.1 GM Survey Meter . . . . . . . . . . . . . . . . .103 8.3.2.2 Nuclear Medicine Thyroid-Uptake Probe . . . . . . . . . . . . . . . . . . . . . . . . . . .104 8.3.2.3 Portal Monitors. . . . . . . . . . . . . . . . . . .105 8.3.2.4 Nuclear Medicine Gamma Camera . . .106 8.3.2.5 Pulse-Height Analyzer and Radionuclide Windows . . . . . . . . . . . . . . . . . . . . . . . .107 8.3.3 Measuring Internal Contamination with Hospital Equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .110 8.3.3.1 Using GM Survey Meters to Assess Internal Contamination with Certain Gamma-Emitting Radionuclides . . . . .110 8.3.3.2 Using a Gamma Camera or ThyroidUptake Probe to Assess Internal Contamination with Certain GammaEmitting Radionuclides . . . . . . . . . . . .110

viii / CONTENTS

8.4

8.5

8.3.4 Wound Monitoring . . . . . . . . . . . . . . . . . . . . . . . Indirect (in vitro) Determination of Internal Contamination . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.4.1 Nasal Swabs . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.4.2 Urine Samples . . . . . . . . . . . . . . . . . . . . . . . . . . 8.4.3 Blood Samples . . . . . . . . . . . . . . . . . . . . . . . . . . 8.4.4 Fecal Samples . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.4.5 Analysis of Indirect Bioassay Samples . . . . . . . Rapid Screening of Persons to Identify Radionuclide Intakes that Exceed the Clinical Decision Guide . . . . .

112 113 113 114 115 116 116 117

9. Medical Management of Internally-Contaminated Persons . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 121 9.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 121 9.2 General Clinical Guidance for the Treatment of Internal Contamination . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 122 9.3 Making Requests for Radiological Countermeasures . . 124 9.4 Medical Management Decisions for PotentiallyContaminated Persons . . . . . . . . . . . . . . . . . . . . . . . . . . 125 9.4.1 Triage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 125 9.4.2 Prioritizing Children and Pregnant Women . . . 126 9.4.3 Medical Decisions During the Early Phases of a Radiation Incident . . . . . . . . . . . . . . . . . . . . . . . 127 9.4.4 Choice of Decorporation Therapy for InternallyDeposited Radionuclides . . . . . . . . . . . . . . . . . . 128 9.4.5 Self-Treatment . . . . . . . . . . . . . . . . . . . . . . . . . . 129 9.4.6 Contaminated Wound Management . . . . . . . . . 129 9.4.7 Using DTPA on Radionuclide-Contaminated Wounds . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 130 9.5 Medical Management of an Americium-241 Intake . . . 131 9.5.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 131 9.5.2 Treatment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 131 9.5.3 Medical Follow-Up After Treatment . . . . . . . . . 134 9.6 Medical Management of a Cesium-137 Intake . . . . . . . 134 9.6.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 134 9.6.2 Treatment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 135 9.6.3 Medical Follow-Up After Treatment . . . . . . . . . 135 9.7 Medical Management of a Cobalt-60 Intake . . . . . . . . . 135 9.7.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 135 9.7.2 Treatment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 138 9.7.3 Medical Follow-Up After Treatment . . . . . . . . . 138 9.8 Medical Management of an Iodine-131 Intake . . . . . . . 138 9.8.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 138 9.8.2 Treatment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 140 9.8.3 Medical Follow-Up After Treatment . . . . . . . . . 140

CONTENTS

9.9

9.10

9.11

9.12

9.13

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Medical Management of an Iridium-192 Intake . . . . . .140 9.9.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .140 9.9.2 Treatment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .145 9.9.3 Medical Follow-Up After Treatment . . . . . . . . . .145 Medical Management of a Plutonium-238 or -239 Intake . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .145 9.10.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .145 9.10.2 Treatment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .146 9.10.3 Medical Follow-Up After Treatment . . . . . . . . . .146 Medical Management of a Radium-226 Intake . . . . . . .146 9.11.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .146 9.11.2 Treatment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .147 9.11.3 Medical Follow-Up After Treatment . . . . . . . . . .147 Medical Management of a Strontium-90 Intake . . . . . .147 9.12.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .147 9.12.2 Treatment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .147 9.12.3 Medical Follow-Up After Treatment . . . . . . . . . .150 Medical Management of a Uranium-235 or -238 Intake 150 9.13.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .150 9.13.2 Treatment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .153 9.13.3 Medical Follow-Up After Treatment . . . . . . . . .153

10. Social, Psychological and Communication Issues Associated with Screening and Monitoring a Population . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .155 10.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .155 10.2 General Approach . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .156 10.3 Program and Center Names . . . . . . . . . . . . . . . . . . . . . .156 10.4 Enlisting the Public as a Partner . . . . . . . . . . . . . . . . . .156 10.5 Communicating with Members of the Public . . . . . . . . .157 10.6 Proactive Approach . . . . . . . . . . . . . . . . . . . . . . . . . . . . .157 10.7 Information Hotlines . . . . . . . . . . . . . . . . . . . . . . . . . . . .158 10.8 Special Services, Approaches and Materials for Children . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .159 10.9 Persons with Reproductive and Fertility Concerns . . . .160 10.10 Other Special Populations . . . . . . . . . . . . . . . . . . . . . . . .161 10.11 Addressing Staff Concerns and Information Needs . . . .161 10.12 Staff Support . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .162 10.13 Training Exercises . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .162 10.14 Understanding and Addressing Responder Concerns and Information Needs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .163 11. Long-Term Follow-Up of Individuals . . . . . . . . . . . . . . . . .165 11.1 Identification of the Population to be Followed . . . . . . .165 11.2 Classification of Persons to be Monitored or Followed . .168 11.3 Follow-Up Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . .171

x / CONTENTS 12. Scalability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 172 13. Assessment of Current Capacity in the United States to Perform Population Screening, Decontamination and Monitoring for Internal Contamination . . . . . . . . . . . . . . 180 13.1 Availability of Equipment and Resources . . . . . . . . . . . 181 13.2 Laboratory Capabilities . . . . . . . . . . . . . . . . . . . . . . . . . 182 13.3 Training Needs on Use of Equipment . . . . . . . . . . . . . . 183 13.4 Radiation Volunteers to Support Population Screening 185 13.5 Biodosimetry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 186 13.6 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 187 14. Conclusions and Recommendations . . . . . . . . . . . . . . . . . 188 14.1 Recommendations for Planning . . . . . . . . . . . . . . . . . . . 188 14.2 Recommendations Related to Screening and Treatment of a Population for Internal Contamination . . . . . . . . . 189 14.3 Recommendations for Additional Work . . . . . . . . . . . . . 190 Appendix A. Radiological Properties of Radionuclides Considered in this Report . . . . . . . . . . . . . . . . . . . . . . . . . . 192 Appendix B. How to Perform a Radiation Survey for Contamination: Instructions for Workers . . . . . . . . . . . . B.1 Screening Survey . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B.2 Complete Whole-Body Survey . . . . . . . . . . . . . . . . . . . . B.3 Most Common Mistakes Made During the Survey . . . .

202 202 203 204

Appendix C. How to Distinguish Between Alpha, Beta and Gamma Radiation Using a Geiger-Muller Survey Meter . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 205 C.1 Determining the Presence of an Alpha-Emitting Radionuclide Using Only a Geiger-Muller Survey Meter . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 206 C.2 Determining the Presence of Strontium-90 (or other pure beta emitters) Using a Pancake Geiger-Muller Survey Meter . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 206 Appendix D. Survey and Registry Forms . . . . . . . . . . . . . . . . 208 D.1 Contamination Survey Sheet . . . . . . . . . . . . . . . . . . . . . 208 D.2 Registry Form . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 211 D.3 ATSDR Rapid Response Registry Survey Form ATSDR Rapid Response Registry Survey Form . . . . . . . . . . . . . 214 Appendix E. How to Perform Decontamination at Home . . 216

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Appendix F. Using Geiger-Muller Survey Meters to Assess Internal Contamination for Selected Gamma-Emitting Radionuclides . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .218 Appendix G. Collection and Preparation of Biological Samples for Radioanalysis . . . . . . . . . . . . . . . . . . . . . . . . . .228 G.1 Urine Samples . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .228 G.2 Main Collection Issues . . . . . . . . . . . . . . . . . . . . . . . . . . .228 Appendix H. Shipping of Biological Samples . . . . . . . . . . . . .231 H.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .231 H.2 Regulatory Information: Brief Summary . . . . . . . . . . . .231 H.3 Definitions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .232 H.4 General Shipping Guidelines for Clinical, Diagnostic and Exempt Biological Samples . . . . . . . . . . . . . . . . . . . . . . .232 H.5 Specific Shipping Guidelines for Clinical, Diagnostic and Exempt Biological Samples . . . . . . . . . . . . . . . . . . . . . . .234 H.6 Website Links for Some Commercial Shippers . . . . . . .234 Appendix I. Population Screening and Monitoring Implications of Two Urban Contamination Incidents . .235 I.1 Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .235 I.2 Overview of Two Cases . . . . . . . . . . . . . . . . . . . . . . . . . .236 I.2.1 Goiânia, Brazil . . . . . . . . . . . . . . . . . . . . . . . . . . .236 I.2.2 London, United Kingdom . . . . . . . . . . . . . . . . . .236 Appendix J. Pregnancy Categories for Drug Use . . . . . . . . . .238 Appendix K. Emergency Phone Numbers for Government Officials to Request Assistance Following a Radiological or Nuclear Incident . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .239 Glossary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .240 Abbreviations and Acronyms . . . . . . . . . . . . . . . . . . . . . . . . . . .247 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .248 The NCRP . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .264 NCRP Publications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .273

1. Executive Summary This Report is the second of two reports by the National Council on Radiation Protection and Measurements (NCRP) that focus on measurement of radionuclides deposited internally in a population exposed in a radiological or nuclear incident. The first report, NCRP Report No. 161, entitled Management of Persons Contaminated with Radionuclides (NCRP, 2008a), is an update and expansion of NCRP Report No. 65, Management of Persons Accidentally Contaminated with Radionuclides (NCRP, 1980) that provides detailed guidance for many radionuclides in a much broader range of exposure scenarios. The present Report focuses on screening a population exposed to one or more radionuclides that may be involved in a radiological or nuclear incident. Screening as used in this Report means rapid assessment and measurement for external or internal contamination. Screening is a singular activity intended to enable intervention and management of persons who may have been or who were exposed to radioactive contamination from a radiological dispersal device (RDD) or a nuclear incident. Thus, a group of people who may be contaminated are first screened for external contamination (a single survey). If contaminated externally, they are decontaminated prior to screening for internal contamination (a single measurement). The emphasis is on rapid screening that quickly identifies those patients who may need medical treatment to decorporate internally-deposited radionuclides. This Report uses the term monitoring in a broad sense, which is either the monitoring of a population for internal contamination by taking a number of measurements or collecting a number of bioassay samples over a period of time or the medical monitoring of a population for subsequent medical effects depending on the context. External monitoring for radioactive contamination is not addressed in detail in this Report but is discussed in much greater detail in other referenced literature. The radionuclides in this Report were selected based on information published by various regulatory and advisory organizations including the Centers for Disease Control and Prevention (CDC, 2009) and the U.S. Department of Health and Human Services (DHHS, 2009). This Report addresses screening a population for internal contamination and the possible use of decorporation 1

2 / 1. EXECUTIVE SUMMARY therapy for patients who exceed a level of internally-deposited radionuclide called the Clinical Decision Guide (CDG), a concept developed in NCRP Report No. 161 (NCRP, 2008a). This Report provides information intended for use by emergency responders and planners and public-health officials for development of emergency response plans that include screening a population for internally-deposited radionuclides. However, this Report is not intended to be a procedure manual that can be incorporated directly into an existing radiological response plan. This Report should be used to evaluate and upgrade, to the extent feasible, the capability to screen small, medium and large populations for the presence of internally-deposited radionuclides and to assist in decisions about the possible medical treatment of patients who contain levels of radionuclides in excess of the CDG. This Report also discusses broadly the monitoring of a population for external contamination and decontamination as necessary prior to screening for internal contamination and refers to other literature that treats these subjects more completely. Background information is provided, including various settings in which members of the public might incur internal depositions of various radionuclides. These sections may be of particular interest to those who are involved in development of emergency response plans. Initial screening of individuals at the scene and hospital and mass screening are described, and previous experience with internal radionuclide contamination of members of the public is summarized. Incident command is described briefly to provide the reader with an appreciation for the importance of incident response coordination between the community and the hospital(s) that will be receiving potentially-contaminated patients. It also emphasizes the importance of timely communications during an incident to optimize medical care and treatment of patients to decorporate internally-deposited radionuclides. Efficient and frequent communications facilitate the ability of hospitals to tailor their response to the number of patients who need treatment of acute injuries and screening of patients for internal contamination. The process of sorting victims from a radiological incident based on their risk of having significant radiation exposure or contamination is referred to as radiological triage. The outcome of radiological triage depends partly on whether people receive direct irradiation or contamination from the plume or the contaminated environment. For purposes of this Report, it is assumed that most people who are impacted by the incident or who perceive that they may be contaminated will wish to be screened for contamination.

1. EXECUTIVE SUMMARY

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The CDG can be used by physicians to consider whether the level of radionuclide intake by a patient warrants use of decorporation therapy. This concept was first published in NCRP Report No. 161 (NCRP, 2008a) to provide guidance on when physicians may want to consider the use of decorporation agents to reduce the radiation dose that a patient might receive from an internally-deposited radionuclide. Hospital equipment discussed in this Report that may be used to determine whether the CDG has been exceeded includes nuclear medicine cameras and uptake probes, portal monitors, multichannel analyzers, and portable survey meters for direct screening. Procedures discussed for indirect screening include use of nasal swabs and the collection and analysis of urine and fecal samples. Physicians will be particularly interested in Section 9, which provides guidance on the medical management of patients who have been identified as containing one or more internally-deposited radionuclides. Section 9 provides information on the use of the CDG in making treatment decisions and general guidance for treating patients. It also provides information on how patient care providers can request equipment, supplies and pharmaceuticals from the Strategic National Stockpile (SNS) (CDC, 2008). Medical management guidance is restricted to the radionuclides of interest in this Report and includes consideration of treatment with decorporation agents and over-the-counter products. In addition to the many technical and logistical issues associated with screening patients for internal contamination; social, psychological and communications issues are important. This Report provides guidance on organizing and conducting a screening program in a manner that is sensitive to these issues. It also offers guidance on practical needs such as information hotlines, setting up population screening centers, and considering the needs of special populations including children and pregnant women. Exposure of a population to radioactive materials from an RDD or an improvised nuclear device (IND) may increase the risk of deterministic or stochastic effects from external exposure and internally-deposited radionuclides. This Report describes a long-term follow-up program to monitor (i.e., to follow the health of) this population. Guidance includes the immediate identification of patients who should be included in long-term follow-up and a survey form that includes appropriate identification information for long-term follow-up. Screening activities can be readily undertaken for a few patients but can become complicated or impractical when large numbers of patients present themselves. This Report discusses scalability of emergency response plans and recommends that communities and

4 / 1. EXECUTIVE SUMMARY hospitals consider dividing their responses into three classes based on the numbers of people involved: small (e.g., 1 to 10 people); medium (e.g., tens of people), and large (e.g., over 100 people). Also included in this Report is an assessment of the current capacity within the nation to perform rapid screening of a population exposed to radioactive materials from a radiological or nuclear incident. Surveys of manufacturers and state radiation-control program directors suggest that most communities that responded to the survey have the necessary resources to provide rapid screening of a limited number of patients to determine whether they are internally contaminated. NCRP (2008a) identified major areas of research and development needed to address the management of contaminated persons. These areas are decontamination facilities to remove external contamination, instrumentation and modeling for assessment of internal contamination, bioassay facilities and automation, biomarkers and devices for biodosimetry, software for rapid estimates of organ equivalent dose and effective dose, decorporation agents for use postexposure, medical follow-up of exposed populations, and educational programs in health physics. In addition, this report identified the need for research to develop a more complete understanding of the capabilities of states, communities and hospitals to respond to a radiological incident and to screen populations of different sizes and in different locations for internal contamination. While most communities probably could provide appropriate screening and treatment of a limited number of internally-contaminated patients, more complete information is needed to fully understand capabilities to respond to incidents that create a need to screen large numbers of patients. Several appendices to this Report provide both practical and more detailed information. While the purpose of this Report is to address internal contamination, accurate screening for internal contamination cannot be performed with survey instruments if the patient has external contamination. The appendices contain practical advice on how to perform radiation surveys, how to distinguish between alpha, beta and gamma radiation using a Geiger-Mueller (GM) survey meter equipped with a pancake probe, and how to perform decontamination at home. Survey and registry forms are provided to enter patients into a registry for recordkeeping and for long-term follow-up. GM count rates from CDC (2009) are provided to determine whether a patient has internal contamination in excess of the appropriate CDG values. Procedures are provided for preparation of urine samples and shipping biological samples for analysis. The appendices contain a summary of the 210Po poisoning of a single individual in United Kingdom and contrasts that with the

1. EXECUTIVE SUMMARY

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large-scale 137Cs contamination incident in Goiânia, Brazil. The purpose for these case studies is to compare the procedures used to identify the radionuclide and the individuals who were exposed and to describe monitoring for external contamination and screening for internal contamination. Emergency phone numbers are provided for requesting government assistance following a radiological or nuclear incident. Finally, various U.S. Food and Drug Administration (FDA) drug categories for pregnant women are defined. This Report makes several recommendations and emphasizes a number of points that are highlighted within the various sections. For preplanning purposes, emergency planners should be aware that radiological and nuclear incidents have the potential to cause widespread contamination of people and the consequences will place enormous stress upon emergency response and healthcare organizations. In the development of an emergency response plan, emergency planners should recruit volunteers such as health physicists, radiation-safety officers, and medical personnel who could serve as subject matter experts or perform specific tasks to support planning and response to a radiological incident. Due to their regular duties, nuclear medicine staff may assist but will not be able to support fully the needs of the emergency department for quick screening of patients during a mass casualty incident. Therefore, other hospital staff should be trained and made available to assist with these activities. The planning process should address capabilities for rapid detection and identification of radionuclides. This is particularly challenging for pure alpha- and beta-emitting radionuclides. Screening a population for internal contamination requires planning and practice. In particular, the use of nuclear medicine cameras and thyroid probes to screen patients following exposure to radionuclide contamination from a radiological or nuclear incident requires advance planning, training and rehearsal. It cannot be implemented ad hoc during an emergency. Local municipalities and public-health agencies should develop procedures for requesting, receiving and distributing SNS assets. This planning should include providing the diagnostic and medical management guidelines to healthcare providers for the use of decorporation agents for internally-deposited radionuclides. Planning for population screening should incorporate provisions for establishing stakeholder advisory boards or other mechanisms for making the public a partner and should take into account psychosocial, behavioral and communication issues likely to arise during the screening of a population. Regarding the management of contaminated patients in all but the most extreme cases, standard precautions provide adequate

6 / 1. EXECUTIVE SUMMARY protection to healthcare workers to prevent secondary radioactive contamination. Patients who have suffered life-threatening injuries should be given medical care immediately, without regard to contamination. The presence of external contamination should almost never delay urgent medical care. Emergency personnel should be aware that patients with the most serious injuries are also likely to be the most contaminated both externally and internally. The presence of internal contamination is rarely life-threatening to a patient. Therefore, treatment of conventional injuries that may be immediately life threatening should take precedence over decorporation of internal contamination. Rapid identification of the radionuclide(s) involved in a radiological or nuclear incident is essential to the selection of appropriate methods for assessment of internal contamination and subsequent treatment decisions. The CDG may be used by physicians as a basis for medical treatment of individuals who have internal radionuclide deposition. The CDG is not intended to instruct physicians on a specific course of action such as administration of decorporation agents. Rather, CDGs are intended as a tool to be used to help a physician determine when radiation exposure may have clinical significance. NCRP recommends a bioassay action-level or benchmark equal to one-half of the CDG value for identifying those individuals for whom continued bioassays should be considered for the purpose of correlating internal contamination with long-term biological effects. Community or hospital emergency response plans should include resource needs to scale the response to the size of the incident. The following recommendations are made in this Report to improve capability of the nation to screen populations for internal contamination following exposure to a radiological incident. Additional surveys of local communities and states should be conducted to assess more accurately their capability to screen populations for internal contamination. Consideration should be given to utilizing strategically located clinical laboratories to analyze bioassay samples for radionuclide identification and concentrations. A limiting factor here is the small number of radiochemists in the United States. Additional research should be conducted to define radiological instrument or bioassay trigger levels that correspond to the CDG for a wider variety of radionuclides such as those identified in NCRP Report No. 161 (NCRP, 2008a), which may be produced by the explosion of an IND.

2. Introduction 2.1 Overview Hospitals, first-responder groups, local and state health departments, and nongovernmental organizations have extensive experience responding to a variety of disasters such as major storms and airplane crashes. Experience with some other kinds of threats is far more limited (e.g., no large-scale radiological or nuclear accidents or terrorism incidents have occurred in the United States). In recognition of the need for an all-hazards approach, planners now include response to radiological incidents in their emergency plans. Many hospitals include in their emergency plans a component for responding to an incident that could result in a few patients who are externally contaminated with radionuclides (e.g., an industrial accident involving one or more radionuclides). Until recent years, these plans anticipated small numbers of patients and focused mostly on external contamination. Currently, emergency response planning has increasingly included possible incidents that could result in tens, hundreds, or even larger numbers of patients with both external and internal radionuclide contamination (HSC, 2009). This necessitates development of a flexible, scalable approach to emergency response, one that will help manage the variety, intensity and duration of an emergency (JCAHO, 2009). Thus, for radiation emergency response, the plan should accommodate the variety of radiation sources that could be involved in the emergency, the intensity and magnitude of an incident, and the duration of the incident. Furthermore, for some scenarios hospitals should not assume that conditions during an incident will allow local agencies to support the hospital’s response. Therefore, hospital emergency plans should be flexible (e.g., using their own security staff to control traffic near the entrance to the emergency department if local law enforcement is occupied at the scene of the incident). NCRP Report No. 161 (NCRP, 2008a) presents detailed information on a broad range of exposure settings, radionuclides, and numbers of persons that might be involved. The focus of this Report is much narrower and is directed primarily to actions associated with a population of people that might be involved in a large-scale radiological incident. Therefore, the list of radionuclides of most interest (Table 2.1) is a small subset of the list in Report No. 161 (NCRP, 7

8 / 2. INTRODUCTION 2008a) and is similar to lists being used by other organizations including CDC (2009) and DHHS (2009). The radionuclides 131I, 235 U and 239Pu are also listed relative to a possible nuclear incident. In the event of an incident involving an IND, the number of detectable radionuclides will significantly exceed those considered in this Report and the radioactive fallout will reach across many jurisdictions, potentially involving multiple states (HSC, 2009). Population screening following a nuclear detonation will be conducted primarily to detect and remove external contamination. Considerably more work is needed to develop a robust, rapid screening program to identify internal contamination of survivors of a nuclear detonation and to make medical treatment decisions in the face of unparalleled medical demands. However, in such an incident, decorporation decisions could be based on internal contamination by 131I and 137 Cs, radionuclides that are included in Table 2.1. One of the first tasks for emergency management following a radiation exposure incident or environmental release of radioactive material is to recognize that radionuclides were involved and to identify the radionuclide(s). The U.S. Environmental Protection Agency (EPA) Radiological Emergency Response Team based in the EPA Office of Radiation and Indoor Air and regional offices, provides advice and responds to emergencies involving releases of radioactive materials by providing environmental measurement and guidance activities (EPA, 2010), but communities and states should be aware that they are responsible for initial assessments because the Radiological Emergency Response Team will not be able to respond immediately. Community or state agencies should also request assistance from the U.S. Department of Energy (DOE) Federal Radiological Monitoring and Assessment Center (FRMAC). FRMAC coordinates federal radiological monitoring and assessment activities with those of local and state agencies (DOE, 2009a). Within several hours of notification, DOE will deploy a team of health-physics specialists from the Radiological Assistance Program and will make a determination of whether to send Phase I of a Consequence Management Response Team consisting of technical and management personnel who should reach the site of the emergency within 6 to 10 h. The Radiological Assistance Program and Consequence Management Response Team Phase I response will be supported by the Consequence Management Home Team consisting of technical and scientific personnel from DOE laboratories (NNSA, 2009). A complete FRMAC for the radiological emergency should be fully operational within 24 to 36 h and will operate under the parameters of the Incident Command System (ICS) (Section 5) to

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TABLE 2.1—Characteristics of radionuclides considered in this Report.a Half-lifeb

Emissions

60

5.27 y

β, γ

90

28.8 y

β

Medical treatment, radioisotopic thermal generators

131 d

8.02 d

β, γ

Medical diagnosis and treatment

137

30.2 y

γ

192

73.8 d

β, γ

226

1,600 y

α, β, γ

Self-luminous products, cancer therapy (no longer used, legacy products remain)

235

7.04 × 108 y

α, β, γ

Nuclear reactor fuel, nuclear weapons

238

4.47 × 109 y

α, β, γ

Anti-tank weapons, radiation shielding

238

87.7 y

α, n

239

24,100 y

α

241

432 y

α, γ

Radionuclide

Co

Sr d

I

Cs d

Ir Ra d

Ud Ud Pu

Pu d Am

Usesc

Industrial radiography, stereotactic radiosurgery, commercial irradiators

Radiography, well logging, soil density gauges, self-shielded irradiators Industrial radiography, brachytherapy

Neutron generators, calibration sources, radioisotopic thermal generators Nuclear weapons Soil moisture, hydrocarbon content, smoke detectors

aAdditional data on the radiological characteristics of the radionuclides listed in this table and their progeny are given in Appendix A. b ICRP Publication 107 (ICRP, 2008). cSee NCRP Report No. 161 (NCRP, 2008a) for typical source activities. d Likely presence of radioactive decay progeny of importance as well as the indicated radionuclide.

10 / 2. INTRODUCTION develop a monitoring and screening plan. The Interagency Modeling and Atmospheric Assessment Center will provide plume model predictions which will be helpful in determining where people were relative to the plume when they became contaminated. But once again, communities should prepare to provide immediate response to the emergency including emergency medical treatment and radiological response. Because federal support may not arrive for hours, local emergency management will be responsible to provide appropriate triage, screening, and subsequent external decontamination of individuals as needed. Local emergency management and health professionals can call the CDC Operations Center for immediate telephone advice on emergency response and emergency patient care. After external decontamination of each casualty, screening for internal contamination should be completed. However, emergency patients should be stabilized before external decontamination is attempted beyond clothing removal. Other literature sources provide guidance and methodology on external decontamination of patients (e.g., CDC, 2007; CRCPD, 2006; NCRP, 2001; 2005; 2008a). Ideally, first responders would survey medically-stable patients for external contamination and decontaminate them as necessary before release or transportation to a hospital. However, in a large-scale incident, some people are likely to self-evacuate and report directly to an emergency department or community reception center. Therefore, medical personnel should confirm that external decontamination is complete prior to screening or treatment of the patient for internal contamination. Emergency response personnel should be aware that the most seriously injured are also the most likely to be externally and internally contaminated due to their proximity to the incident (DHHS, 2009). After confirmation that external decontamination is complete, an assessment of potential internal contamination should be made (Section 8). The goal of screening the exposed population is to determine whether internal contamination is high enough to justify appropriate medical treatment to expedite biological removal of the internal contamination (decorporation therapy) thereby reducing the subsequent irradiation of critical tissues (Section 9). Clinical decisions on decorporation therapy are guided by a CDG for each radionuclide. The CDG concept is discussed in Section 7. The CDG represents a level of internal radionuclide deposition at, or above, which a physician may wish to consider medical treatment to enhance removal of the radionuclide from the body. Members of the public may be identified for bioassay, medical treatment, biodosimetry, dose reconstruction, and entry into long-

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term follow-up programs and registries (Section 11). The purpose of screening the population is to reduce radiation exposure to people by making timely and adequate measurements of potentiallycontaminated individuals so they can receive prompt and adequate external decontamination, medical treatment of injuries, and decorporation therapy. The current capacity to implement these activities at the local and state levels is limited (Section 12). However, the local community should be prepared to monitor for external contamination, decontaminate, screen for internal contamination, administer pharmaceutical countermeasures (decorporation), and enter patients into a registry during the first 24 to 48 h or more after which federal assistance should become available. Section 13 provides an assessment of the current capacity within the nation to perform rapid screening of a population exposed to radioactive materials from an RDD, IND, or similar mass casualty incident. Surveys of manufacturers and state radiation-control program directors suggest that most communities have the necessary resources to provide rapid screening of a limited number of patients to determine whether they are internally contaminated. Resources considered include those discussed in Section 8. Accurate screening for internal contamination will be difficult in patients with external contamination. This Report contains several appendices that provide both practical and more detailed information on screening. Appendix A provides additional information on the radiological properties of radionuclides covered in this Report. Appendix B explains how to perform screening and wholebody surveys for external contamination and Appendix C provides information on how to distinguish between alpha, beta and gamma radiations using a GM survey meter. Related survey and registry forms are given in Appendix D. Appendix E contains instructions for people who perform self-decontamination of external radionuclide contamination at home. Appendix F provides important information on how GM survey instruments can be used to assess internal depositions of certain gamma-emitting radionuclides. Detailed instructions for the collection of bioassay samples for in vitro bioassay analyses are given in Appendix G and instructions for possible shipments of these samples to analytical facilities are given in Appendix H. Appendix I provides a summary of the 210Po poisoning of a single individual in London, United Kingdom and contrasts that with the large-scale 137Cs contamination incident in Goiânia, Brazil. The purpose of Appendix I is to compare the procedures used to identify the radionuclide and the individuals who were exposed and to describe screening and monitoring of individuals. Appendix J defines FDA drug categories for pregnant women

12 / 2. INTRODUCTION and Appendix K contains emergency phone numbers for contacting government officials to request assistance with a radiological or nuclear incident. NCRP has a long history of providing recommendations on response to radiological emergencies and has issued several publications that offer specific advice on response to a radiological emergency. The first of these reports, NCRP Report No. 29, Exposure to Radiation in an Emergency (NCRP, 1962), provided general advice and was superseded by NCRP Report No. 42, Radiological Factors Affecting Decision-Making in a Nuclear Attack (NCRP, 1974), which provided guidance for use in response to a large-scale nuclear disaster involving an intense and uncontrolled exposure of a population. NCRP Report No. 65, Management of Persons Accidentally Contaminated with Radionuclides (NCRP, 1980) was a comprehensive report directed primarily to occupational contamination incidents occurring in nuclear facilities. Many data from, and ideas on, treatment of contaminated patients were collected in this report and its recommendations were directed toward physicians responsible for managing these patients. Report No. 65 provided treatment information for many chemical elements and described pertinent information on numerous radionuclides. NCRP Commentary No. 10, Advising the Public About Radiation Emergencies: A Document for Public Comment (NCRP, 1994) focused on communication issues. NCRP Report No. 138, Management of Terrorist Events Involving Radioactive Material (NCRP, 2001), dealt with releases of radioactive materials as the result of a deliberate act of terrorism and provided recommendations on medical management of contaminated individuals, psychosocial issues, communication with members of the public and media, and training of responders. Commentary No. 19, Key Elements of Preparing Emergency Responders for Nuclear and Radiological Terrorism (NCRP, 2005) focused on detection and personal protection equipment requirements for emergency responders, decontamination equipment and medical supplies, and radiationprotection training of emergency responders. NCRP Report No. 156, Development of a Biokinetic Model for Radionuclide-Contaminated Wounds and Procedures for Their Assessment, Dosimetry and Treatment (NCRP, 2006) provides advice on treatment of individuals who may have been contaminated internally through wounds. The most recent report, NCRP Report No. 161, Management of Persons Contaminated with Radionuclides (NCRP, 2008a) is an update of NCRP Report No. 65 that provides recommendations for a broader range of radionuclides and exposure scenarios and for onsite and prehospital actions when responding to contamination incidents and the management of contaminated patients at the hospital.

2.2 PURPOSE OF THIS REPORT

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CDC is the lead federal agency charged with assisting local, tribal and state agencies with population monitoring and decontamination (CDC, 2005a). CDC hosted a Population Monitoring Roundtable in January 2005 that convened representatives from various federal agencies, local and state public-health agencies, clinicians, private public-health organizations, and health professional organizations to assist CDC in evaluating the best methods and techniques for performing internal population screening for radiation, inventorying current U.S. capability to perform this screening, and developing guidance that local and state publichealth agencies can use in preparing to respond to a radiological or nuclear incident (CDC, 2005b). Roundtable participants recognized the complexity of determining internal contamination in a population that had been contaminated externally. They did not identify specific methodologies for determining internal contamination but suggested that external contamination might be a good indicator for internal contamination, and that scalability should be considered. They also recommended that brief histories be elicited for exposed or potentially-exposed individuals, to include basic demographic information and a description of the circumstances regarding their exposures. This information could be used subsequently if resources are available to conduct follow-up for late health effects. They also identified challenges in treating patients for internal contamination including identifying those who need treatment, addressing scalability for the number of patients who are identified as containing internally-deposited radionuclides, and specific agebased guidelines for treatment. 2.2 Purpose of this Report The purpose of this Report is to provide recommendations for local, state and federal responding communities regarding internal radionuclide depositions, decorporation therapy, equipment and laboratory needs. This Report is intended to provide general advice on screening members of the public for internal contamination following an emergency during which people may have been contaminated with radionuclides. Any early radiation injury will most likely be due to external radiation exposure from large particles that fall quickly to the ground and are not respirable. Internal contamination is unlikely to pose a short-term danger to the population (NA/IOM, 2009). Nevertheless, it is important to be aware that following the explosion of an RDD or IND, patients with the most serious medical injuries are likely to have been close to the explosion and are likely to be internally contaminated. Also, in the first

14 / 2. INTRODUCTION 10 to 15 min, anyone within 500 m of the explosion is at risk of inhaling a large amount of activity (Harper et al., 2007). This Report also provides advice on treatment methods to reduce internal radionuclide depositions in individuals who exceed a CDG of radioactive material. Finally, this Report provides recommendations for registering contaminated individuals in a registry for longterm follow-up of internally-contaminated patients. 2.3 Target Audiences of this Report This Report should be useful to a wide range of preparedness and response organizations, emergency management jurisdictions, and policy makers at the local, state and federal levels involved in emergency preparedness and response. However, this Report is aimed primarily at emergency responders including hazardous materials teams, hospitals, and health departments (NCRP, 2005) who will be involved in screening for external contamination followed by decontamination as necessary, treatment of physical injuries, screening for internal contamination, and providing decorporation therapy, if necessary, in patients who may have been exposed to radionuclides during an incident. This Report is based largely on recommendations in other publications, particularly NCRP Report No. 161 (NCRP, 2008a) and is intended as a guide to those who would participate in the screening of small to large numbers of patients who may contain internal radionuclides from exposure to radioactive contamination following a radiological or nuclear incident. This Report is presented in sections that address separate issues including screening the exposed population, treatment of those who have quantities of an internally-deposited radionuclide that warrant removal efforts, and long-term follow-up of the exposed population. Thus, emergency responders may find the sections on screening a population to be most useful whereas physicians, nurses, and healthcare providers may find most value in the section on treatment of patients. However, the entire Report will be useful to those who are responsible for emergency planning. 2.4 Scope of this Report This Report addresses screening of a population, small or large, that may have been exposed to radioactive material from a radiological incident. This population includes emergency responders and anyone potentially exposed to radioactive materials from the incident. It does not address screening of workers exposed to radionuclides as a result of other types of employment, members of the public exposed to patients who have been administered radioactive

2.4 SCOPE OF THIS REPORT

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drugs or sources, or any other situation addressed by a licensed program that uses radioactive material. The radionuclides specifically covered in this Report (Table 2.1) have been limited to those that CDC (2009) has highlighted plus 226 Ra. NCRP (2008a) discusses numerous additional radionuclides and provides more detailed internal dose estimates. Bioassay methods in use today were developed and designed for the adult worker (e.g., whole-body counter size and metabolic models). Unlike adult workers, young people and members of the public who have special needs require unique methods, models and protocols to screen for the presence of radionuclides and to estimate radiation dose. This Report discusses and provides advice on the following issues related to screening and treatment of members of the public of any age who are internally contaminated: • recommended methods, protocols and practices that enable determination of internal contamination: - direct bioassay (in vivo); and - indirect bioassay (in vitro such as urine, feces, or nasal swabs). • elimination of external contamination prior to direct and indirect bioassays; • social, psychological and communication issues related to population monitoring and screening; • radionuclide-specific baseline values measured in the U.S. population, or if none, whether a national baseline data gathering program should be recommended; • age-specific internal contamination trigger levels (CDG values) that would indicate that decorporation therapy should be considered [e.g., countermeasures such as Prussian blue or diethylenetriamine pentaacetic acid (DTPA)]; • recommendations for the receipt, physical security, distribution, and administration of SNS assets; • recommendations for maintaining medical and pharmaceutical supplies and assets locally; • age-specific internal contamination trigger levels that indicate the need for a person to be entered in a long-term follow-up program or registry, and guidance for the following: - administration of a long-term health surveillance program; - content of the personal history needed to plan for subsequent long-term follow-up (i.e., specific questions to ask); and - type of follow-up testing needed.

16 / 2. INTRODUCTION • recommendations against the use of self-administered decorporation therapy (e.g., laxative, aluminum and magnesium antacid, etc.); • discussion of scalability of the above recommendations based on the size and magnitude of the incident or potentiallyexposed population; and • discussion of the current capacity in the United States to perform these functions, what is needed to augment these capacities, the training recommendations and to whom this training should be provided.

3. Background Information Radioactive materials are used in many types of settings, and people may become contaminated or may experience an intake of radionuclides in nearly all of these settings. Knowledge of the setting in which a contamination incident has occurred can help determine many relevant factors surrounding the incident, including the type of radiation involved (alpha, beta or gamma), the likelihood of an intake, the type of intake (inhalation, ingestion, absorption through the skin, or through wounds), the chemistry and biokinetics of the radionuclides, and the number of people likely to be involved. However, while large numbers of people in a densely populated area near the detonation of an RDD or IND might become contaminated and require external decontamination, few if any will be internally contaminated to a level that requires medical treatment (DHHS, 2009). After intake, radionuclides are distributed within the body according to the physical and chemical properties of the materials in which they are incorporated or the elemental chemical properties of the radionuclides after dissolution in body fluids. Some radionuclides are preferentially deposited in a particular organ or tissue and impart a higher radiation dose to it than to the remainder of the body, whereas others are distributed relatively uniformly in the body, imparting similar doses to most tissues. Until they are excreted from the body or diminished by radioactive decay, internally-deposited radionuclides will irradiate body organs and tissues. Radionuclides have different radiological properties [e.g., halflife, type(s) of radiation emitted, and specific activity]. The chemical form of a radionuclide, which influences its distribution within the body, and the radiological properties determine the radiation doses received by various organs and tissues. Detailed discussions of dosimetric aspects of the radionuclides covered in this Report can be found in NCRP Report No. 161 (NCRP, 2008a). 3.1 Internal Deposition of Radionuclides Skin provides a barrier that keeps most potentially-harmful contaminants outside the body. Exceptions include tritiated water and unbound radioiodine. However, contaminants can enter the body if the skin is breached or if they are inhaled or ingested (Figure 3.1). 17

18 / 3. BACKGROUND INFORMATION

Fig. 3.1. Generic biokinetic diagram showing routes of entry, metabolic pathways, and possible bioassay samples for internally-deposited radionuclides (adapted from ICRP, 1997).

3.1.1

Inhalation

The inhalation of radionuclides can lead to health risks to the persons exposed. Particles of certain sizes may penetrate deeply into the lungs, come into contact with body fluids, dissolve and enter the systemic circulation. Less soluble particles may remain in the lungs for months or years chronically irradiating the lung tissue and providing a longer-term opportunity for dissolution of the particles and absorption of their contents into the blood. Increased knowledge of anatomy and physiology of the respiratory tract and of the behavior and biological effects of inhaled radioactive particles is reflected in the detailed models published by the International Commission on Radiological Protection (ICRP) in Publication 66 (ICRP, 1994a) and Supporting Guidance 3 (ICRP, 2002a). These models make it possible to calculate radiation doses to various regions of the respiratory tract from the inhalation of airborne particles for a wide range of exposure conditions. NCRP Report No. 125 (NCRP, 1997) also provides a summary of scientific information and mathematical models that describe the deposition and clearance of various inhaled radionuclides from studies of exposed laboratory animals and people.

3.1 INTERNAL DEPOSITION OF RADIONUCLIDES

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Factors that influence the radiation dose pattern received from inhaled radionuclides (ICRP, 1994a) include: • Particle size: The aerodynamic and thermodynamic characteristics of inhaled radioactive particles determine whether particles are of an inhalable size, whether they are deposited or exhaled, and if deposited, where in the respiratory tract (ICRP, 1994a). Large (i.e., >10 μm) and fine (<0.01 μm) particles may only reach the nasal passages because of impaction and diffusion processes, respectively, whereas particles of intermediate sizes may reach the deepest and smallest airways. Relatively-insoluble particles that are deposited on the ciliated portions of the respiratory tract are cleared by mucociliary activity up and out of the respiratory tract, swallowed, and passed through the gastrointestinal (GI) tract or are expectorated. The clearance of particles deposited deeper in the lung, in the nonciliated alveolar-interstitial region, is particularly influenced by their solubility in body fluids. • Solubility: The ICRP lung model (ICRP, 1994a) classifies the solubility of inhaled radioactive particles into three absorption types: F (fast), M (moderate), and S (slow). Type F particles dissolve quickly and their contents are absorbed into the systemic circulation followed by radionuclide translocation and deposition in various body organs and tissues and excretion in the urine and feces. At the other end of the solubility spectrum, relatively-insoluble, Type S, particles dissolve very slowly and they are retained in the deep lung for much longer periods of time, causing the lungs to receive a larger radiation dose than other body tissues or organs. Particles with Type M solubility demonstrate intermediate characteristics of both pulmonary retention and dissolution, absorption or translocation to other body organs and tissues. • Mode of radioactive decay: The committed equivalent dose to the lungs depends on a number of factors including the type(s) of radiation emitted by the deposited radionuclide, total number of radionuclide transformations in the lungs, energy absorbed per unit mass of lung tissue, and the radiation weighting factor(s) of the emitted radiation(s). The fractional energy absorption of gamma radiation emitted in the lung is less than that of alpha and beta radiations, which are almost entirely absorbed in the lung. Alpha radiation is more damaging than beta or gamma radiation to the living cells of the lungs and it has a radiation weighting factor of

20 / 3. BACKGROUND INFORMATION 20 compared to one for both beta and gamma radiations. In addition, many alpha emitters give rise to a chain of radioactive progeny, which can deliver radiation dose over years or decades. Both external and internal contamination by gamma-emitting radionuclides can be assessed by counting with external detectors. Measurements of external contamination by alpha- and beta-emitting radionuclides are moderately easy using alpha and GM detectors, respectively. Because alpha and beta radiation from internal contamination cannot penetrate the chest wall, screening for internally-deposited contamination is often indirect, including taking nasal swabs, sampling excreta, or determining the amount of radioactive materials in the air to which a person was exposed. 3.1.2

Ingestion

Radionuclides may enter the GI tract either directly or indirectly (ICRP, 2006). Direct ingestion occurs when contaminated food or water is ingested. Indirect ingestion occurs when radioactive particles suspended in the air are inhaled and deposited on the lining of the mouth, nasal cavity, and throat and subsequently swallowed. In either case, the radioactive material enters the GI tract where it may be absorbed into the body, or excreted, or both to some extent. Factors that affect the radiation dose received from ingested radioactive materials include the: • chemical form and solubility of the radioactive material; • nature of the ingestion (e.g., the presence of food in the digestive tract may affect absorption of the radioactive material and may provide more shielding against alpha or beta radiation than liquids or mucus alone); • type(s) of radiation(s) emitted during radioactive decay (alpha, beta or gamma); • residence time in the digestive tract (if, for example, trauma or shock causes the digestive tract to slow, radioactive materials may produce a higher radiation dose to the digestive organs); and • biokinetics of the ingested radionuclides absorbed into the systemic circulation. Rapid assessment of internal contamination due to ingestion of radioactive materials is also important. As with inhaled radioactive materials, only gamma-emitting radionuclides are easily detected

3.1 INTERNAL DEPOSITION OF RADIONUCLIDES

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from outside the body. If the ingested radionuclide emits only alpha or beta radiation, sampling of excreta may be the only way to confirm ingestion of radioactive material. As such, if time and resources do not permit evaluation by bioassay, it may be appropriate to assume that a person has ingested radioactive materials and to take actions accordingly until an assessment can be made. 3.1.3

Absorption from Skin Contamination

Although the skin will normally act as a barrier to keep radioactive materials on the outside of the body, some chemicals are absorbed through the skin. Because of this, skin contamination can sometimes lead to the intake of radioactive materials. Radionuclides of iodine and tritium, for example, are easily absorbed through the intact skin but most other radionuclides are not. In the case of skin contamination, an external survey will usually suffice to identify the extent of contamination for all forms of radioactive material. 3.1.4

Absorption Through Wounds

Radionuclides may be introduced directly into the bloodstream via open wounds, by the embedding into the body of shrapnel that is contaminated with radioactive particles, or by the embedding of radioactive particles during any incident that breaches the skin. These scenarios include the possibility of radionuclide uptake into the systemic circulation resulting in radiation doses to internal organs (NCRP, 2006). Factors that can affect radiation dose from the introduction of radioactive materials via wounds include: • chemical form and solubility of the radioactive materials; • type of wound (e.g., puncture wound that retains foreign material as opposed to open wound in which bleeding cleans the wound); • depth of penetration of the radionuclide(s) into the body; • size of the injected particles (e.g., large particles dissolve more slowly into the bloodstream than do smaller particles with similar chemical characteristics); and • biokinetics of the radioactive materials reaching the systemic circulation. When present in wounds, radioactive materials, unlike those that are inhaled or ingested, can sometimes be detected from outside the body (even when the radioactive materials emit alpha or beta radiation) because some radioactive materials may be detectable on the skin, within the wound, or in smears of blood from the

22 / 3. BACKGROUND INFORMATION wound (NCRP, 2006). In addition, other assessment methodologies such as excreta sampling or screening for internal gamma emitters may be appropriate. Dose from a wound uptake may be higher than the dose from a similar amount of radioactive material inhaled or ingested because the biological mechanisms that limit uptake, such as mucociliary clearance and GI absorption, are bypassed. A wound can provide a direct path for material in a soluble form to enter the blood after dissolution and absorption. 3.2 External Contamination In most cases, external contamination of the skin is far more likely to occur than is an intake of radioactive materials. However, skin contamination can lead to an intake if contaminated skin comes in contact with food or directly with the mouth causing ingestion, or a person brushes his or her nose with a contaminated hand resulting in some inhalation exposure. For example, a person under stress may bite his or her fingernails, a hungry person may touch food with contaminated hands, and a child may suck his or her fingers. In addition, a few radioactive materials (e.g., radionuclides of iodine) may be in a chemical form that is absorbed through the skin leading to internal contamination. Although skin contamination does not normally lead to high radiation exposure, skin contamination with beta emitters may lead to severe localized beta burns (Barabanova and Osanov, 1990). Due to the short range of most beta particles in tissue, the beta particle energy is deposited in a relatively small thickness of skin and can cause localized burns if not removed quickly (NCRP, 1999; 2008a). Some radionuclides, such as 32P and 90Sr/90Y, emit some beta particles of sufficient energy to penetrate the skin and irradiate underlying tissues. 3.3 Effects of Weather Weather can have a significant effect on an incident that involves radiological or nuclear material. Rain can wash radioactive materials from the air, reducing the distance that contamination can travel. This also concentrates the activity into a smaller area that may result in a higher radiation exposure to the people in that area. In the case of a dispersal of radioactive materials, the effects of weather will likely prove even more significant. In a chemical or nuclear explosion, radioactive materials are violently thrown into the air and contamination will spread to some distance from the scene regardless of the weather. By comparison, dispersal of radioactive materials may result in distribution in the atmosphere

3.4 COMPLICATIONS

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where they are likely to persist for a longer period of time prior to detection. Such an incident will likely be more sensitive to ambient weather conditions. Snowfall, rain, fog, and high humidity are often sufficient to suppress dust formation, or to bring dust back to Earth quickly. Under such conditions, contamination from the dispersal of radioactive materials may be less widely spread, concentrating the activity in a smaller area with concomitantly higher radiation dose rates. However, if the contamination is not cleaned up, it may be mobilized later, when the weather changes. In general, it is possible to make predictions about the effects of weather on the spread of contamination from a radionuclide release as shown in Table 3.1. Another meteorological condition that may affect the spread of contamination from a radionuclide release is a temperature inversion. Temperature inversions help trap airborne materials near the ground. When this happens, light surface winds (temperature inversions are typically associated with calm surface weather conditions) may suspend low quantities of contamination, but they will remain close to the surface. This may reduce the area contaminated while increasing the radiation dose and the amount of inhaled radioactive materials among those exposed to the more concentrated radionuclides in the trapped air (Eisenbud and Gesell, 1997). Large cities may experience yet another effect that may impact the spread of contamination from a radiological or nuclear incident (i.e., the “urban canyon” effect). This refers to the manner in which long ranks of tall buildings can channel the wind, effectively directing the spread of contaminants. This can cause contamination to spread a greater distance along the streets owing to higher wind velocities. This may affect the amount of contamination deposited, radiation dose rate to those in the affected areas, and the amount of airborne radioactive materials that might be inhaled or ingested. 3.4 Complications Due to the Presence of Multiple Agents or Serious Injuries Hospitals can expect little or no warning before casualties begin arriving from the scene of a mass casualty incident. In addition, information regarding the hazardous agent(s) is not likely to be available immediately. Consequently, The Joint Commission requires hospitals to develop an all-hazard approach to be flexible enough to respond to emergencies of all types (JCAHO, 2009). In a best practices document, the Occupational Safety and Health Administration provided practical information to help hospitals address employee protection and training for mass casualty incidents involving a variety of hazardous substances (OSHA, 2005).

Dry

Humidity

Limited to immediate area and a meter or less of altitude

Tens to hundreds of meters downwind and into breathing zone

Hundreds of meters or more downwind and tens of meters above ground

Calm

Breezy

Windy

Wind

Horizontal and Vertical Spread of Contamination

Those downwind in the vicinity of the release may be exposed to high levels of airborne radionuclides, but high or gusting winds may dilute activity to lower levels further away

Radioactive material concentrations may be harmful to those in the vicinity of the device and up to several tens of meters away

Radioactive material concentrations will be highest close to ground, may be harmful to those on or near the ground (depending on the amount of radioactivity released)

At Time of Initial Release

Moderate to high levels of resuspension at scene of release and nearby

Low to moderate levels of resuspension near ground level in vicinity of release

Very limited resuspension, limited spread of radioactive materials following release

Resuspension of Contamination

Impact of Weather Condition on Contamination Spreada

TABLE 3.1—Impact of weather conditions on the dispersal or resuspension and effects of radionuclide releases (AEC, 1968; Connell and Church, 1980; Eisenbud and Gesell, 1997; Hosker, 1980; OTA, 1979).

24 / 3. BACKGROUND INFORMATION

Radioactive materials likely to remain near site of release, increasing initial radiation and contamination levels

Radioactive materials released from device may wash away up to hundreds of meters in heavy rain, unlikely to rise to breathing zone Contamination likely to remain near site of incident due to relatively low winds usually associated with temperature inversions

Calm Breezy Windy

Usually calm to very light winds

Wet (rain, snow)

Temperature inversion

Temperature inversions are usually not long-lived; it is likely that conditions will change in a period of a few to several hours

Resuspension unlikely except in vicinity of fires

Possible resuspension in the vicinity of highest contamination levels or in presence of fires

Little resuspension expected

A release of radioactive materials may take place over an extended period of time, possibly reducing the impact of relatively transitory meteorological conditions on the release.

a

Breathing zone radioactive materials concentrations are unlikely to pose a risk

Many tens to hundreds of meters downwind, likely to reach several meters to tens of meters into air

Windy

Radioactive material concentrations may rise to high levels for those in the vicinity of the release

Tens of meters downwind, may or may not extend to breathing zone

Breezy

Radioactive material concentrations are not likely to be harmful

Limited to immediate area, not likely to extend to breathing zone

Calm

Damp (fog, mist)

3.4 COMPLICATIONS

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26 / 3. BACKGROUND INFORMATION The presence of multiple harmful agents also may complicate the treatment of patients (Hall, 2000). For example, caustic substances may damage the skin, enhancing the entry of radioactive materials into the bloodstream. Similarly, inhaling chemical agents may damage the lung tissue, speeding absorption of the radioactive materials, or increase mucus production which could enhance mechanical clearance or isolate the epithelium more effectively, leading to decreased absorption. Alternately, the presence of acids or bases may change the chemical form of inhaled radioactive materials, possibly enhancing or inhibiting absorption through the lungs or through the skin. Other complicating factors include serious injuries. Although it is often tempting to first attend to radiological matters, serious medical conditions (especially those that are life-threatening) require immediate attention. An injured contaminated person or patient should never be permitted to die or to suffer needlessly because of undue attention to radiological assessment or decontamination. In most cases, radiological hazards are unlikely to be lifethreatening, although there have been fatalities resulting from radioactive materials inhalation or ingestion [e.g., the Goiânia 137Cs incident (IAEA, 1998)]. Accordingly, potentially-serious medical concerns should take precedence over radiological concerns in most cases. Only if the patient is stable should full attention be given to radiological issues. 3.5 Radiological Triage Radiological triage is the process of sorting people involved in a radiological incident based on their risk of having significant radiation dose or contamination (DOHMH, 2009). The radiological triage procedures used will depend partly on the nature of the radiation exposure (e.g., direct irradiation versus contamination from the plume) or the contaminated environment. Information on location of individuals will be a prime indicator of direct irradiation (Rojas-Palma et al., 2009) versus contamination by the plume (Harper et al., 2007). For purposes of this Report, it is assumed that most people who are impacted by the incident or who perceive that they may be contaminated will wish to be screened for contamination. Section 6 provides general guidance on radiological triage and screening individuals for contamination. 3.6 Proximity to the Incident After an explosive dispersal of radioactive materials, the probability of severe injury and likelihood of radioactive contamination

3.7 PREVIOUS EXPERIENCE WITH INTERNAL CONTAMINATION

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through wounds or inhalation will be affected by the proximity of people to the explosion. In addition, people who were not injured but were present 10 to 15 min after the explosion and within 500 m of the explosion are at risk of internal contamination. Even people who were protected from the explosion by being inside a building could be exposed to significant radioactivity if they passed through the plume after exiting the building (Harper et al., 2007; Musolino and Harper, 2006). People with the most serious injuries are also likely to be contaminated both externally and internally.

Emergency responders should expect injured people to be contaminated externally and should decontaminate and transport them accordingly. Patients with life-threatening injuries should be transported immediately without decontamination by wrapping them in sheets to prevent contamination of ambulances. At the hospital, these patients should be stabilized prior to decontamination. After decontamination, they should be screened for internal contamination. While most people exposed to the radioactive materials from the detonation of an RDD or IND might become contaminated and require decontamination, few, if any, will be contaminated to a level that requires medical treatment (DHHS, 2009). However if, in the physician’s judgment, the patient is likely to be highly contaminated internally (e.g., severe injuries indicate that the patient was close to the blast), the physician should consider administration of decorporation agents as discussed in Section 9. 3.7 Previous Experience with Internal Contamination Historically, a number of incidents have occurred involving varying degrees of internal contamination by all of the pathways noted above. Although most such incidents have involved a relatively small number of contaminated people, several incidents contaminated large numbers of people. These include the 1986 Chernobyl nuclear reactor accident (Balonov, 2007), the 1987 incident in Goiânia, Brazil (IAEA, 1988), and several incidents during the era of atmospheric nuclear weapons testing (e.g., Gilbert et al., 2002; Robison and Hamilton, 2010; Simon and Bouville, 2002). Additional details are provided in Section 4.

28 / 3. BACKGROUND INFORMATION 3.8 Conclusions A radiological or nuclear incident may lead to the external and internal contamination of large numbers of people. In addition, there have been radiological contamination incidents (e.g., Chernobyl; Goiânia, Brazil), some resulting in many cases of external and internal contamination. Radiological factors influencing the severity of a contamination incident include the radionuclide(s) involved, their decay mode(s) and energies, the mode of exposure (e.g., inhalation, ingestion, injection), the physical and chemical form(s) of the radioactive material involved, the amount of radioactive material, and the number of people exposed. In a large-scale contamination incident, external factors such as the weather, location and mode of dispersal, the terrain, and the presence of other agents may exacerbate the effects of the incident. At the same time, the reactions of many involved in any such incident may further complicate the incident response. An appropriate response to any large-scale radiological contamination incident should, at a minimum, account for these factors.

4. Settings in Which Persons May Become Contaminated with Radioactive Material Identification and evaluation of the types of contamination incidents that may occur is an important part of emergency preparedness. This enables emergency responders and medical personnel to incorporate appropriate treatment of contaminated individuals into emergency plans. NCRP Report No. 161 (NCRP, 2008a) discussed over a dozen settings in which individuals may become contaminated with radionuclides. Of these, only a few settings are likely to contaminate a large number of people and thus are relevant to this Report. These settings are reviewed briefly in this section. While not a large-scale incident, the 2006 London 210Po poisoning is instructive and relevant (Harrison et al., 2007). The mass amount of 210Po administered was miniscule, but led to the contamination of several aircraft, a number of buildings, and at least 50 people, in addition to the death of the person who was targeted. The spread of contamination reportedly resulted from an attack on a single person, using a very small amount of radioactive material; it is likely that an assault against a population using larger amounts of radioactive materials will lead to a wider spread of contamination and a larger number of contaminated individuals than was seen in London. Examples can be found in Karam (2005), Mettler and Voelz (2002), and Musolino and Harper (2006). 4.1 Radiological Dispersal Device There has been speculation about the nefarious use of radioactive materials in a radiological dispersal device (RDD), commonly referred to as a “dirty bomb.” An RDD uses conventional explosives or some other mechanism to spread radioactive contamination (NCRP, 2005). When incorporated into an explosive device and detonated, radioactive materials could be spread over a large area by the explosion. These radioactive materials might then contaminate people ranging from a handful of individuals close to the blast up to tens, hundreds, or even thousands of people, depending on the 29

30 / 4. SETTINGS IN WHICH PERSONS MAY BECOME CONTAMINATED nature of the incident. In addition to this external contamination, the airborne radioactive material might be inhaled, exposing those affected to internal radionuclide deposition as well. 4.1.1

Incident Characteristics

For purposes of emergency preparedness it is assumed that the dispersal of radioactive materials following an RDD incident will be somewhat similar to the dispersal associated with various radiological incidents that have resulted in contamination of the environment or members of the public. For example, the radiological contamination incident in Goiânia, Brazil (IAEA, 1988; Oliveira et al., 1991) serves as an example of dispersal of radioactive material from an incident (Appendix I). Lessons learned from such contamination incidents can be used in the development of various elements of emergency plans such as rapid screening and decontamination of those affected. Some inferences that may be important in developing an emergency plan are noted below. • Persons closest to the explosion are likely to be both seriously injured and most highly contaminated. These people are also the most likely to experience contaminated wounds, and to inhale radioactive materials following the explosion. • Fears of radiation and contamination may lead to widespread concerns among many who are not directly affected by the explosion. This may lead, in turn, to hospitals, aid stations, and related facilities being overwhelmed by those seeking assistance. • Fears of radiation exposure and its effects on the part of medical providers and emergency responders may impact the timeliness of emergency and medical response efforts. These fears may be alleviated by proper training of these individuals. • Administrative policies and practices (e.g., removal of outer clothing, designated decontamination areas) on the management of contaminated individuals may be used to help reduce contamination levels and radiation dose. 4.1.2

Nature of Contamination

People at the scene are likely to experience moderate to high levels of contamination on their clothing and skin, may inhale or ingest radioactive materials, and may have contaminated wounds. People downwind of the site of the explosion may have low to moderate levels of skin and clothing contamination, and may have low

4.2 AEROSOL DISPERSAL INTO A PUBLIC AREA

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levels of internal contamination from inhalation or ingestion. People upwind of the site of the explosion are unlikely to be contaminated unless the incident takes place in an urban environment. While large numbers of people in a densely populated area around the detonation of an RDD might become contaminated externally and require external decontamination, few if any will be contaminated internally to a level that requires medical treatment with decorporation agents (DHHS, 2009). 4.2 Aerosol Dispersal into a Public Area An RDD is likely to spread radioactive contamination into the air, making it available for inhalation. However, a less obvious incident such as a nonexplosive dispersion event could introduce radioactive materials into the atmosphere in aerosol form. Such an incident may go unrecognized for some time, possibly exposing a larger number of people to inhaled radioactive materials. 4.2.1

Incident Characteristics

By its nature, aerosol dispersal might be recognized only after the contamination had affected many people and a large area. If the quantities of radioactive materials used were small, there may be few or even no injuries even if contamination is widely spread. However, large levels of external contamination may be capable of causing radiation injury by direct contact with the source material, such as was observed in the Goiânia, Brazil 137Cs incident (IAEA, 1998). However, exposure to dangerously high levels of radioactive materials by ingestion or inhalation is not likely. An aerosol dispersal incident might be detected by chance, as contamination spread to research universities, hospitals, or nuclear power plants is discovered during the performance of routine radiological surveys. Once detected, determining the extent to which the contamination had spread, identifying people who were contaminated and retracing the pathway to the origin of the radiological dispersal incident may prove a significant challenge. In some cities, routine radiological screening or the use of radiation pagers (small hand-held or belt clip-on radiation detectors) by law enforcement and emergency responders may also lead to the detection of an aerosol dispersal incident. 4.2.2

Nature of Contamination

Radioactive contamination from an incident of this nature may consist of a variety of particle sizes that may be in many chemical forms. Those affected may inhale or ingest radioactive materials,

32 / 4. SETTINGS IN WHICH PERSONS MAY BECOME CONTAMINATED but the very nature of such an incident will most likely preclude “injection” via radioactive shrapnel. If the aerosol is water-soluble (e.g., cesium chlorine), it may dissolve into rain, snow, or other atmospheric moisture (e.g., mist, fog); otherwise, water may be adsorbed on the particles, making them settle to the ground more rapidly. Since an incident of this nature would not likely be recognized immediately, it is possible that a great deal of time might elapse before it is recognized; this could allow wider dissemination of the radioactive materials before the incident is discovered. It would also allow more time for biological elimination and physical decay of internally-deposited radionuclides, which would complicate interpretation of screening results. 4.3 Contamination of Food or Water Supplies Food and water supplies could become contaminated following an accident involving radioactive materials, and some have speculated the deliberate and surreptitious introduction of radioactive contamination into food or water supplies (Allan and Leitner, 2006). Though either incident may cause widespread fear and cause many to seek medical attention and radiological screening, it is not likely that these incidents would cause any fatalities or acute illnesses because the method of distribution would likely dilute the radioactive materials to the point of relative harmlessness. In addition, water treatment plants are quite effective at removing unwanted contaminants (Brinck et al., 1976; Stetar et al., 1993), further reducing the efficacy of this method of dissemination. Similarly, radioactive contamination of the food supply may harm a few people (by contamination close to the point of consumption) or may have a lesser impact on many (by contamination at the point of production). Thus, it may be possible to harm a few people, or to contaminate large numbers of people. In light of the above, contamination of food and water supplies could lead to extensive contamination of a population resulting in the need to rapidly assess and (potentially) decontaminate large numbers of people. 4.3.1

Incident Characteristics

While an explosive RDD incident would be quickly recognized and the radiological component (presumably) identified a short time later, contaminated food or water might not be recognized for days or longer. In fact, such an incident may go unrecognized, barring accidental discovery in the course of routine radiological surveys or a public announcement by the group responsible. However, this delay may serve to heighten public fears, and would certainly

4.4 IMPROVISED NUCLEAR DEVICE

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serve to greatly increase the number of the people who might consider themselves contaminated. 4.3.2

Nature of Contamination

Contamination of food and water supplies would almost certainly lead to internal contamination with possibly a combination of alpha-, beta-, or gamma-emitting radionuclides. It may also lead to widespread external contamination as people handle contaminated food or wash and cook with contaminated water. This contamination may be widespread, depending on the type and amount of radioactive material involved and the length of time between initiation and discovery of the incident. 4.4 Improvised Nuclear Device An improvised nuclear device (IND) refers to any type of explosive device designed to produce a nuclear explosion (NCRP, 2005). Use of an IND would be likely to cause widespread destruction, serious damage to infrastructure in the vicinity of the attack, extensive radioactive contamination, and an overwhelming mass casualty scenario in which medical care would be austere (NA/ IOM, 2009). The nuclear bombings of Hiroshima and Nagasaki serve as examples of the effects of an IND, although an IND may have significant differences as described below. Additional details of the anatomy of a nuclear detonation can be found in the National Academies/Institute of Medicine workshop report (NA/IOM, 2009). 4.4.1

Incident Characteristics

The nuclear bombings in Japan in 1945 caused widespread damage to the cities’ infrastructure, including destruction of many buildings, severance of utility lines, and mass fires (Glasstone and Dolan, 1977). An IND is expected to have similar characteristics, with some crucial differences. Most of these are a result of a presumed difference in detonation altitude. In particular, the nuclear weapons used at the end of World War II were detonated at an altitude designed to maximize the blast damage from the device. By so doing, the fireball did not contact the ground, which led to relatively low levels of radioactive fallout. However, a surface burst (the presumed altitude of an IND) would pull vaporized and irradiated earth up into the fireball resulting in high levels of radioactive fallout and concomitantly elevated levels of surface radiation after the detonation. Post-detonation radiation and contamination levels are likely to be higher than were noted in either Hiroshima or

34 / 4. SETTINGS IN WHICH PERSONS MAY BECOME CONTAMINATED Nagasaki. Within minutes of, and for several hours after, the detonation and for kilometers downwind, fallout would pose immediate danger to life and health of emergency responders as well as anyone who is outdoors for even short periods of time (NA/IOM, 2009). Radiological concerns in the aftermath of an IND may include mild to lethal radiation sickness, heavily contaminated persons (potentially with contaminated wounds or other medical concerns), and potentially high radiation levels from residual fallout in the vicinity of the detonation. Accordingly, it will be necessary to evaluate patients for radiation injury in addition to physical injury and other medical concerns (HSC, 2009). Patients can receive a dose of radiation from fallout in three ways: • directly from external radiation as the fallout passes by or from the fallout that has been deposited on the ground; • from fallout contamination on skin, clothing or possessions, which exposes people until they change their clothing or remove the contaminated material; and • ingestion or inhalation of radioactive material. Of these, the most likely to cause injury in the first few days is direct exposure to fallout (DHHS, 2009; NA/IOM, 2009). An IND is also likely to cause many nonradiological effects that may complicate emergency response efforts. Electromagnetic pulse is one of these effects; a nuclear detonation generates a substantial electromagnetic field that will induce electrical currents in conductive materials. These currents can destroy sensitive electronic devices, including radios, computers, and related devices. Because of this, communications may be impaired or destroyed, vehicles (many of which rely on computers for proper operation) may fail to operate properly, computers (and related devices such as radios, cellular phones, personal digital assistants, bar code scanners, etc.) may be rendered inoperable, and a city’s electrical grid may shut down. Any or all of these factors will serve to complicate emergency response efforts. Another nonradiological effect that will complicate emergency response is flash blindness, a temporary blindness caused by the brilliant flash of light associated with the detonation. Flash blindness is likely to last from several seconds to several minutes incapacitating anyone in which sudden loss of vision could cause a traffic accident or other untoward incident (NA/IOM, 2009). In addition, the destruction caused by an IND may destroy much of a city’s infrastructure through blast damage, crater formation, and the initiation of mass fires (sometimes referred to as “firestorms”); further hindering response and recovery efforts. The

4.5 NUCLEAR REACTOR INCIDENT

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destruction of roads, water mains, natural gas lines, electrical systems, and other utilities is likely to delay all aspects of emergency response, including all phases of medical assessment and care (HSC, 2009). This may necessitate transporting injured persons to locations where an intact infrastructure will allow proper treatment and care. 4.4.2

Nature of Contamination

Radioactive contamination in the aftermath of a nuclear detonation will include a large number of radioactive fission products such as radioisotopes of iodine, cesium, technetium and rubidium. It is also likely to contain neutron activation products formed from the neutron irradiation of the weapon casing, buildings and structural materials, and the soil. Finally, there will likely be low levels of unfissioned uranium or plutonium, along with minor amounts of transuranic elements formed by neutron capture by unfissioned uranium and plutonium. In short, the scene of the IND and those areas affected by fallout, as well as people (both injured and uninjured) in these areas will likely be contaminated with alpha-, beta-, and gamma-emitting radionuclides. Depending on weather conditions, the radioactive fallout may extend for many kilometers from the site of the attack, or may remain relatively confined to the area(s) near the explosion. Radioactive materials may be inhaled, ingested, embedded in the body as shrapnel and absorbed through breached skin, and may be present as contamination on the skin or clothing. 4.5 Nuclear Reactor Incident The accident at the Chernobyl nuclear reactor spread detectable levels of radioactive materials over much of Europe and led to the evacuation and relocation of over 345,000 residents in nearby areas (Balonov, 2007). However, the risk of an incident of similar severity occurring in a Western-designed nuclear power plant is very low because of differences in design. These include the inherent stability characteristics of the reactor design and the use of multiple layers of containment, including a reinforced containment structure surrounding the entire reactor plant. These same containment structures also reduce the likelihood of an external explosion causing a large-scale release of radioactive materials to the environment. The 1979 incident at the Three Mile Island Nuclear Power Plant resulted in the almost complete destruction of one reactor’s core, and is perhaps a more appropriate example of what might occur in the event of an incident at a Western-designed nuclear reactor. Although the Three Mile Island incident received a great deal of

36 / 4. SETTINGS IN WHICH PERSONS MAY BECOME CONTAMINATED attention from the media and the public, the average radiation dose to those members of the public living within 16 km of the plant was <0.1 mSv (10 mrem) (Fabrikant, 1981). The actual off-site radioactive releases from that accident were primarily short-lived noble gases that did not pose an environmental or personnel contamination threat. In March of 2011, an earthquake followed by a tsunami caused a loss of electric power from the grid and backup sources of power at the Fukushima Daiichi Nuclear Power Plant in Japan. Subsequently, above normal levels of various radionuclides were measured in air and water emissions from these power plants. At the time of printing of this Report, final measurements of radioactivity concentrations in the environment were not available. However, Becker (2011) pointed out the immense psychosocial footprint of such a disaster including people suffering anxiety disorders, depression, a persistent sense of ill health, and multiple unexplained physical symptoms. Over the years emergency preparedness activities near nuclear power plants have helped local and state planners develop plans that will be of use in responding to RDD events that require screening the population for external and internal contamination. 4.5.1

Incident Characteristics

A large-scale contamination incident at a nuclear power plant could lead to the contamination of workers and buildings at the nuclear power plant. It is also possible that personnel could be exposed to sufficiently high radiation levels to cause radiation injury or death; although there have been few such accidents in the history of nuclear reactors.1 The probability that a Western-designed nuclear power plant will release large amounts of radioactive materials to the environment is low, but the possibility must be considered, especially in light of the problems experienced at the Fukushima Daiichi Nuclear Power Plant (Becker, 2011). If this occurs, there may be widespread contamination. It should also be noted that a commercial nuclear power plant cannot undergo a nuclear detonation because the uranium fuel used contains <10 % of the fissile radionuclide, 235U; these concentrations of 235U are too low to produce a nuclear explosion. 1These include the 1961 incident at the U.S. Army’s SL-1 reactor (Horan and Gammill, 1963) and a 1958 incident at a Yugoslavian research reactor (Hurst et al., 1961).

4.6 LARGE-SCALE FIRES AND INCIDENTS

4.5.2

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Nature of Contamination

Contamination from a nuclear-reactor incident could consist of fission and activation products. People inside the plant may inhale or ingest radionuclides, and they are likely to have contaminated skin and clothing. However, radioactive shrapnel is not likely to be present. Relatively short-lived noble gases are the most likely environmental release from an operating nuclear reactor, with little or no potential for personnel contamination. The radionuclide likely to produce the greatest radiation dose to the population is 131I. The population that is within the 10 mile (~16 km) emergency planning zone is at greatest risk of contamination. The U.S. Nuclear Regulatory Commission (NRC) has supplied potassium iodide (KI) tablets to states that have requested them for distribution in the incident of an emergency. KI could be used to supplement evacuation or protection in place. When the population is evacuated out of the area, the risk of additional 131I exposure to the thyroid gland is essentially eliminated. Beyond 10 mile (~16 km), the major risk of radioiodine exposure is from ingestion of contaminated foodstuffs, particularly milk products. NRC requires emergency preparedness programs for nuclear power plants, which are designed to protect the public against contamination from nuclear power plant incidents. 4.6 Large-Scale Fires and Incidents Several nuclear facilities have experienced large-scale fires that have spread, or threatened to spread radioactive materials. These include wildfires at the Los Alamos National Laboratory (Kraig et al., 2001), the 1957 and 1969 plutonium fires at the Rocky Flats Plant (Mongan et al., 1996; Rood et al., 2002), the Windscale reactor fire (Crick and Linsley, 1984), and the explosion at the Soviet Union’s Mayak Nuclear Weapons Facility in 1957 (Goldman, 1997). The Los Alamos wildfires involved relatively small levels of contamination present in the soils, the Rocky Flats plutonium fires occurred in an industrial facility, and the Mayak explosion involved the release of large amounts of radioactive materials (hundreds of thousands of gigabecquerel) over hundreds of square kilometers. 4.6.1

Incident Characteristics

It is reasonable to expect that any large-scale incident such as a fire or explosion has the potential to release large quantities of radioactive materials and to spread them over large areas. If such an incident occurs in an unpopulated location, the numbers of people exposed may be quite small, perhaps limited to emergency

38 / 4. SETTINGS IN WHICH PERSONS MAY BECOME CONTAMINATED response and on-duty personnel. However, should such an incident occur in a highly-populated area (such as was the case with Mayak and Chernobyl), thousands to hundreds of thousands of people may be contaminated internally, externally, or both. Such numbers will place considerable strain on available resources. 4.6.2

Nature of Contamination

Contamination released by a fire or explosion will initially be airborne, resulting in external contamination, inhalation, and potential ingestion risks. The radionuclides involved will be those present at the scene, depending on the nature of work occurring at the facility (or site) of the incident. It is reasonable to assume that particles of all sizes may be emitted by an explosion, although the largest particles will likely settle to the ground quickly, and in the vicinity of the incident. However, it is likely that the particles created by a fire would be dependent on the properties of materials in the fire and on the mass of material in the cloud (Parkhurst and Guilmette, 2009), smoke-sized particles from burning radioactive materials, and slightly larger (but still small) particles suspended by the fire itself or resuspended by fire-induced winds. 4.7 Sealed Radioactive Source Incidents Numerous incidents of lost radioactive sources have resulted in human injury (IAEA, 2003). The radiological hazard associated with sealed radioactive sources depends on the activity of the source, the type of radiation emitted, distance from the source, and time of exposure. An intact sealed source will not cause radioactive contamination. However, the dose from direct exposure to an intact sealed source can be extremely high and cause life-threatening whole-body doses and acute injury to skin or other body parts. Accidental exposures to lost industrial radiography sources have been the most common form of radiation incident, and have resulted in serious radiation burns, amputations, and deaths (Gusev et al., 2001). Several incidents in which sealed radioactive sources have been damaged have led to a release of radioactive materials. These include the incident in Goiânia, Brazil (IAEA, 1988; 1998) and the accidental melting of a radioactive source at the Acerinox Facility in Los Barrios, Spain in 1998 (IAEA, 2005a). Sealed sources may contain many terabecquerel of activity, and the release of this radioactive material may cause large-scale contamination, including internal contamination. Such contamination caused four deaths in the Goiânia incident.

4.7 SEALED RADIOACTIVE SOURCE INCIDENTS

4.7.1

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Incident Characteristics

Radioactive sources vary greatly in the activity they contain, from less than a kilobecquerel to many terabecquerel. This makes the effects of a breached source difficult to predict. In most cases, a leaking radioactive source is a nuisance, but is rarely life-endangering; although a high-activity source such as the one involved in the Goiânia incident can be deadly. In other instances, such as the Los Barrios incident, there may be minor amounts of internal contamination, possibly causing no fatalities at all (IAEA, 2005a). In spite of variability in the characteristics of sealed sources, there are some likely commonalities among such incidents: • Radiation sickness and serious radiation burns can result from contact with or close proximity exposure to an intact high-level sealed source. • Radioactive contamination resulting from the inadvertent breach of a sealed source may initially go unrecognized. Because of this, radioactive contamination may become widely spread before the incident becomes known. • Although large numbers of persons may be externally contaminated with radioactive materials, it is not likely that more than a few will become internally contaminated to the extent of causing observable deterministic health effects (Section 9). 4.7.2

Nature of Contamination

The radioactive materials contained within sealed sources have a variety of physical forms including metallic alloy, ceramic, powder, liquid or gas. However, for emergency planning purposes sealed sources should be considered dispersible. The amount of radioactive materials contained within sealed sources can differ by many orders of magnitude. However, in spite of this variety, some general statements are appropriate: • Metallic sources (e.g., ribbons, pellets or wires of 192Ir) are unlikely to spread contamination because solid metals are not easily dispersible. • Gaseous sources (e.g., 85Kr) are most likely to disperse into the atmosphere and are unlikely to cause contamination, with the possible exception of gas atoms or molecules that might be adsorbed onto surfaces or that might lodge in pores or other surface irregularities.

40 / 4. SETTINGS IN WHICH PERSONS MAY BECOME CONTAMINATED • Ceramic sources (e.g., 90Sr) may chip or spall, possibly creating small quantities of dispersible contamination. • The most significant levels of contamination are likely to originate from radioactive sealed sources in which the radioactive materials are in a powdered or a liquid form. 4.8 Summary While only a few types of radiological incidents such as RDDs and INDs have the potential to cause widespread contamination and to contaminate large numbers of people, these incidents are potentially severe and can place enormous stress upon emergency response and healthcare organizations. Such organizations must be prepared to assess external and internal contamination by any of a number of alpha-, beta- and gamma-emitting radionuclides. In addition to anticipating large numbers of contaminated individuals, hospitals and emergency screening centers should also expect that large numbers of concerned persons will want to be screened for contamination, as happened at the incident in Goiânia, Brazil. For these reasons, it is vital that accurate information about the incident and about the risks (or lack thereof) be conveyed to emergency responders, medical professionals, and to members of the public.

5. Coordination with the Incident Command System 5.1 Introduction The purpose of this section is to provide a general overview of the emergency management system, not to provide recommendations on how emergency management should be organized. Those who are interested (e.g., radiation-safety personnel) in assisting during emergency response should become familiar with the organization of emergency management in their community. In a radiological incident, people trained in radiation safety will be needed to assist with many aspects of the emergency response. It is not uncommon for volunteers to make their way to the scene of any incident to offer their assistance, and people trained in radiation safety may offer their assistance in a radiological incident. It is essential that these volunteers understand that their participation may not be possible unless they have already been incorporated into the Incident Command System (ICS). Therefore, those volunteers with radiological training should contact their local emergency management organization, preferably before a radiological or nuclear incident occurs, to request that they be listed as resources. The following terminology is often used in the context of emergency preparedness and response and is defined here to standardize the definitions within the framework of this section. These terms also appear in the glossary along with other terms relevant to the subject of this Report. • Catastrophic incident: “Any natural or man-made incident, including terrorism, that results in extraordinary levels of mass casualties, damage, or disruption severely affecting the population, infrastructure, environment, economy, national morale, and/or government functions” (FEMA, 2009). • Disaster: An incident that overwhelms normal operations. Critical functions and infrastructure are unable to respond much less return to normal on their own, thus requiring regional or federal assistance (Farmer, 2006). 41

42 / 5. COORDINATION WITH THE INCIDENT COMMAND SYSTEM • Emergency: A sudden, urgent, usually unforeseen occurrence or occasion requiring immediate action. Under the Robert T. Stafford Disaster Relief and Emergency Assistance Act, an emergency means any occasion or instance for which, in the determination of the President, federal assistance is needed to supplement local and state efforts and capabilities to save lives and to protect property and public health and safety, or to lessen or avert the threat of a catastrophe in any part of the United States (FEMA, 2009). • Emergency management: The process to achieve a full state of readiness. There are four phases to every disaster: preparedness, response, mitigation and recovery. Each phase takes planning, time and resources from a wide array of partners to ensure a coordinated response. The primary mission of emergency management is to ensure the entity, institution and staff are prepared to respond to foreseeable disasters (Farmer, 2006). • Federal Emergency Management Agency (FEMA): FEMA is an agency of the U.S. Department of Homeland Security (DHS) and is charged with reducing the loss of life and property and protecting the nation from all hazards, including natural disasters, acts of terrorism, and other man-made disasters, by leading and supporting the nation in a risk-based, comprehensive emergency management system of preparedness, protection, response, recovery and mitigation (FEMA, 2009). • Mitigation: Elimination of the threat or vulnerability or at least lessening the consequences or severity of the disaster (Farmer, 2006). • Preparedness: The process of anticipating potential vulnerabilities and developing comprehensive plans with all involved agencies, educating personnel on the plans, and practicing the plans (Farmer, 2006). • Recovery: Returning to normal, or redefining normal, which is typically the most difficult and of longest duration (Farmer, 2006). • Response: How entities, institutions and people react to the threat (Farmer, 2006). The ICS is an emergency management system that can be adapted to specific incidents, organizations, and available resources (DHS, 2004; EMSA, 2006). A short description is included here to acquaint the reader briefly with its organization and to emphasize that organizational emergency preparedness for radiological incidents, including hospital preparedness, should include a link to the

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community emergency operations center (EOC). In a major radiological or nuclear incident, the EOC would be activated and would control emergency response. If the incident requires resources beyond the capabilities of the community and outside assistance (e.g., state or federal) is desired, the EOC incident commander must request it through the local government. The ICS is designed to provide a response effort that can be scaled according to the magnitude of the incident, including coordination of response by agencies and specialties, as appropriate. This includes the capability to add resources when necessary, as well as releasing resources that are no longer required. Many healthcare organizations use the Hospital Incident Command System (HICS) to respond to emergencies. Important functions of both the ICS and the HICS are to provide a: • clear and predictable chain of command; • clear job description and expectations for each specified position; • set of prioritization checklists; • common language for discussing the incident response; and • link between ICS and HICS. An RDD or IND incident will cause the ICS to be activated. However, an unknown dispersal of radioactive materials may result in a more diffuse incident with contamination and injured persons spread over more than one governmental jurisdiction. In addition, the contaminated persons may not be recognized for days or weeks after the incident. If the site of such an incident can be identified, then the ICS may be appropriate for addressing the problems of controlling that site. However, by the time the site is identified, it may be only a small subset of the entire affected area, resulting in activation of a regional or state-wide coordinated ICS. The following descriptions of the ICS and HICS are not intended to be comprehensive but, rather, to describe briefly some of the important features of these systems. For a more complete explanation, the reader should refer to other resources such as DHS (2004; 2008), EMSA (2006), and FEMA (2008a; 2008b). 5.2 Incident Command System The ICS is a proven system, adopted in the United States, Canada, and the United Kingdom, for managing response to a wide variety of incidents with differing degrees of complexity. In the event of an RDD or IND, the ICS will undoubtedly be activated. However, because there are so few radiological incidents that require activating the ICS, the current community ICS may not include

44 / 5. COORDINATION WITH THE INCIDENT COMMAND SYSTEM radiation-safety personnel and may lack provisions for incorporating radiation-safety considerations into existing procedures. In the development of a response plan for a radiological or nuclear incident, community leadership should identify personnel such as health physicists, radiation-safety officers, and medical personnel who could serve as subject matter experts or perform specific tasks to support planning and response to a radiological or nuclear incident. The ICS structure comprises key positions including incident commander, three command staff (public information, liaison, and safety officers) and four general staff (operations, logistics, planning, and financial chiefs). Additional positions can be added to manage specific incidents (e.g., a medical officer can be added to the command staff) to advise the incident commander on issues regarding medical response to population exposure from a specific agent such as a radionuclide. The incident commander bears primary responsibility for coordinating activities at the scene of an incident. Large incidents may require addition of several command staff positions, including a content expert for radiological incidents, who will report to and advise the incident commander. As the incident progresses, the incident commander may delegate duties to command staff directing ongoing operations, planning the progress of the response efforts, organizing logistics, liaising with other organizations, communicating with the public and the media, and overseeing safety during response and recovery efforts. In the aftermath of an RDD or IND, it is likely that all of these positions will be required. Participants within the ICS are assigned their roles, usually in advance of an incident. Individuals capable of providing specialized technical assistance to the incident commander or those in the field must be designated as having an official function. People who attempt to provide assistance without being so designated may be removed from the scene or even arrested. Further information on the ICS and its operations and organization may be found in many places, including DHS (2004; 2008). Individuals interested in and capable of providing specialized technical assistance to the incident commander or to those in the field should contact their local emergency management organization before a radiological or nuclear incident occurs to request they be listed as resources in the event of a radiological or nuclear emergency and to participate in planning and training for appropriate response to the incident.

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Many cities and counties are likely to have limited equipment and personnel resources in radiation safety and health physics (Section 13). It is important that the community ICS and HICS develop and maintain an inventory of available resources that can be called upon in the event of a radiological or nuclear incident. Potential sources of personnel and equipment resources include, but are not limited to: • local colleges or universities (especially large research institutions); • local nuclear facilities, such as nuclear power plants, DOE facilities, and U.S. Department of Defense facilities; • hospitals (especially large hospitals with nuclear medicine or radiation oncology programs); • industries that use radiation or radioactive materials (including industrial radiographers); • professional organizations such as American Association of Physicists in Medicine, American Nuclear Society, American Society of Radiation Oncologists, Health Physics Society, and Society of Nuclear Medicine, including local chapters of each of these organizations; and • some state radiological control programs maintain lists of certified radiation experts, such as authorized medical physicists or health physicists. Local health physicists who wish to participate in local emergency response beyond assistance within their own organization should volunteer themselves and any equipment that may be made available to local emergency responders to incorporate their availability into the local emergency response plan. Each health physicist may have multiple organizations requesting assistance in a radiological or nuclear incident. These organizations may include the health physicist’s employer, local hospitals, emergency response organizations, the ICS, and even the media. In a radiological or nuclear incident, it may be impossible to honor all of these commitments simultaneously, and attempting to do so may detract from the overall response efforts. Each health physicist has a responsibility to prioritize these commitments (possibly after consultation with his or her employer) so that the most important tasks are accomplished satisfactorily. Before volunteering or committing oneself in a radiological or nuclear incident, individual health physicists should determine what role, if any, their employer expects them to have in an incident. For example, a hospital health physicist is likely to be required to work within the HICS and may

46 / 5. COORDINATION WITH THE INCIDENT COMMAND SYSTEM be unavailable for any other duties, while an industrial health physicist may be free to volunteer to assist in the community emergency response. Similarly, each organization that employs health physicists or that maintains radiation survey equipment should first determine its internal needs in the event of an off-site radiological or nuclear incident. It should then determine what resources may be released to the ICS or HICS. Finally, this information should be conveyed to the local hospital(s) and emergency management agencies, after which the organization may wish to ask for volunteers to assist with a radiological or nuclear emergency response from among its radiation-safety personnel. Such a process should help to meet both community and facility needs, while also ensuring that health physicists arriving at the scene of a radiological or nuclear incident have an official status and appropriate credentials. To summarize the above: • Individuals capable of assisting in the community response to a radiological or nuclear incident should be identified in advance and granted official status to respond to an incident. Simply appearing at the scene and attempting to assist may result in an arrest. • Individual health physicists who would like to assist in the event of a radiological or nuclear incident should coordinate their assistance through their employer. • Organizations that wish to contribute personnel or equipment resources to the ICS or HICS following a radiological or nuclear incident must arrange to do so in advance of an emergency. • Community and hospital emergency management should consult with local organizations, including professional societies, to identify their personnel and equipment resource needs and to meet these needs through local resources whenever possible. • All of these actions should be taken in advance of any incident, and all individuals and organizations involved must periodically reconfirm their needs, resources and willingness to participate in the event of an incident. 5.3 Hospital Incident Command System Many incidents, especially major incidents, involve injuries, both to personnel at the scene at the time of the incident, as well as to emergency responders engaged in the incident response. Many

5.3 HOSPITAL INCIDENT COMMAND SYSTEM

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incidents (e.g., earthquakes, major fires) involve large numbers of patients, which can place a tremendous stress on healthcare organizations. In addition, an incident that involves chemical, radiological, biological, or nuclear agents will pose further complications as hospitals are forced to balance their mission to care for the injured against their desire to avoid contamination. For this reason, many hospitals utilize the Hospital Incident Command System (HICS) to direct their response to disasters or incidents that threaten their ability to adequately care for patients during a major incident (EMSA, 2006). Like the ICS, the HICS describes various positions and their responsibilities. Not only should designated hospital staff be trained to act as HICS staff, but they should know how to integrate their response efforts with those of the EOC and with the regional healthcare system. For example, as soon as a radiological or nuclear incident is identified, the local emergency managers should notify local hospitals of the approximate number of radioactively contaminated patients so that the hospitals can prepare for reception and decontamination, as necessary and appropriate. Similarly, hospitals should communicate their readiness and ability to receive patients back to the EOC. In addition, hospitals may request additional resources through the ICS structure. Hospitals maintain emergency management plans that comprehensively describe their approach to emergencies that could occur in the hospital or community. Those hospitals that could receive radiological patients should include the following preparations in their disaster or emergency response plans: • develop procedures for screening injured persons who may be radiologically contaminated; • develop a communication plan to alert the incident command of the presence of radionuclide-contaminated persons shortly after the contamination is discovered (Section 5.5) or to confirm the presence of contamination that had been previously identified by community emergency response personnel; • establish decontamination criteria (i.e., provide a level of decontamination that a person’s physical injuries and medical condition will permit); • maintain survey meters and decontamination equipment; • designate decontamination areas; • establish and train decontamination teams; • obtain medical supplies specific for triage and treatment of victims of radiological incidents;

48 / 5. COORDINATION WITH THE INCIDENT COMMAND SYSTEM • develop procedures for assessing the amount of external and internal contamination; and • develop procedures for handling radioactively contaminated decedents as described by NCRP (2008a). 5.4 Coordination Between Incident Command System and Hospital Incident Command System Any RDD or IND incident will produce radioactively contaminated patients suffering from physical injuries (e.g., broken bones, embedded shrapnel, burns, etc.). These patients may also have been exposed to internal contamination via inhalation or the presence of embedded radioactive fragments driven by the force of the explosion. While it may be possible to require that all contaminated persons be taken or directed to designated community reception centers or hospitals prepared to accept contaminated patients, it is likely that injured contaminated persons will be taken to the first, or to the nearest available hospital. This will require good communications between the EOC emergency managers and the appropriate members of the HICS staff. In addition, it is important that hospitals communicate back to the community incident commander. some of the information that should be communicated is summarized below. Also, identification of the radionuclide(s) involved in an incident will probably be made on the basis of measurements made at the scene and should be communicated to the hospital as soon as possible. 5.5 Communicating Information from the Scene to the Hospitals and from the Hospitals to the Scene Direct communications may be difficult in the aftermath of any significant incident; telephone circuits may be taxed to their capacity, and radio channels may be filled by emergency responders communicating to each other. In the event of an IND, the electromagnetic pulse may destroy any sensitive electronics that are activated at the time of an incident. However, other channels of communications may remain open (e.g., hand-written messages delivered in person). Whatever mechanism is used to convey information among emergency responders at the scene, public-health officials, and hospitals, it is important that this information be transmitted as detailed in Table 5.1. As one example, if a hospital discovers high levels of alpha activity on a victim’s nasal swabs, this information should be communicated to the scene as quickly as possible so the incident commander can ensure the use of respiratory protection to minimize the risk of inhaling an alpha-emitting radionuclide. Sim-

• Help plan for arrival of patient. • Understand nature of radiological concerns (e.g., beta burns, alpha inhalation, whole-body gamma exposure). • Determine radionuclide biokinetics, assessment and decorporation methodologies. Provide estimate of potentially-contaminated persons to inform medical staff and support request for SNS radiological medical countermeasures. • Plan for receipt of patients (e.g., “hot” entrance). • Obtain appropriate decorporation agents (if applicable). Understand nature and magnitude of expected injuries. Helps indicate whether or not radiation sickness or radiation injury is a concern. Useful in planning further decontamination at hospital (for less-injured patients). Helps with triage; seriously injured patients should receive medical care regardless of extent of radiological contamination, while less-injured patients may withstand light decontamination.

Nature of contamination (internal, external)

Nature of incident (suicide bomb, vehicle bomb, etc.)

Highest measured radiation dose rate at scene

Decontamination measures taken at scene (e.g., removal of outer clothing, shower, none, etc.)

Types of injuries (broken bones, burns, amputations, coronary emergencies, etc.)

Reason

Radionuclide-contaminated persons involved (including alpha, beta or gamma emitting)

Information to be Communicated

About the Incident, from Incident Commander to Hospital Incident Command

TABLE 5.1—Communication between the community and hospital incident commands.

5.5 COMMUNICATING INFORMATION FROM THE SCENE

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Assist with triage and treatment of each victim. Allows targeted decontamination or containment of contamination while serious injuries are being attended to. Useful in planning further decontamination for each victim. If information is available, helps prioritize concerns for each victim. If information is available, can help physicians plan treatment. Helps decontamination workers with conduct of decontamination, helps medical caregivers understand where to give injections, run IV lines. Help medical caregivers appropriately address all of the victim’s concerns. Help determine proximity to blast and resulting radiation dose. Can be used to help determine severity of injuries or radiation exposure (e.g., vomiting at time of rescue may indicate a high radiation dose to victim).

Location of heaviest contamination

Decontamination or other protective measures taken at scene

Medical (e.g., nonradiological, such as broken bones) concerns for each patient

Estimates (if any) of intake

Location(s) of contamination on victim (survey map)

Other considerations (e.g., allergies, psychological distress, etc.)

Location of victim at time of injury

Condition at time of rescue

Reason

Estimated radiation dose to whole body

Information to be Communicated

About Each Patient, from Ambulance Team to Emergency Department Staff (if possible)

TABLE 5.1—(continued)

50 / 5. COORDINATION WITH THE INCIDENT COMMAND SYSTEM

May help incident commander and public-health officials to understand the nature of the incident or attack. Helps identify protective measures that must be taken by those on the scene. Can indicate that emergency responders must take additional precautions; may indicate that existing precautions are adequate. An indication of the overall severity of the incident. May limit quality of care, may restrict types of patients that can be received. If hospital is overwhelmed, may need to request assistance from incident commander. Can incident commander continue sending patients to this hospital?

Specific radionuclide(s) involved (if not already identified at the scene)

Presence of radiation injury (e.g., skin burns, radiation sickness)

Status of patients (e.g., the numbers of dead on arrival, critical condition, good condition, etc.)

Infrastructure impairments at the hospital (e.g., loss of power, loss of water or sewer, etc.)

Problems onsite (e.g., self-referring patients, need for security assistance)

Availability of resources

Reason

Nature and extent of contamination (e.g., heavy external with light internal, moderate external and internal, heavy external, no internal, etc.)

Information to be Communicated

About the Patients, from Hospital Incident Command to the Incident Commander

5.5 COMMUNICATING INFORMATION FROM THE SCENE

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52 / 5. COORDINATION WITH THE INCIDENT COMMAND SYSTEM ilarly, if the incident commander becomes aware of injured people in a radiation area, this information should be communicated to the hospitals giving them time to prepare for the patients’ arrival. Also, identification of the radionuclide(s) involved in an incident will probably be made on the basis of measurements made at the scene and should be communicated to the hospital as soon as possible. In short, communication in both directions is important and can help both the incident commander and the hospitals to better direct their efforts and protect their people. In addition, this communication can help both the medical and emergency response personnel to better anticipate, address and respond to the incident. Communication of information must take place as necessary and appropriate, and it should follow established lines. For example, information should flow from the hospital and from field locations to the EOC emergency managers, who will pass on relevant information to local and regional officials. They, in turn, will pass information and requests on to state and federal agencies and officials.

6. Radiological Triage and Screening Guidance Fundamental to disaster response is the process to sort the injured and to allocate resources in a manner that maximizes the number of survivors (DOHMH, 2009). The scope of triage is broader for radiological and nuclear incidents and includes a group of actions termed radiological triage in a triage, monitoring and treatment handbook by Rojas-Palma et al. (2009). The highest priority following a radiological or nuclear incident is to care for people who have been critically injured (CDC, 2007). This care should be delivered regardless of the level of internal or external radioactive contamination of the injured. Ideally, and subsequent to ensuring that the needs of the critically injured will be addressed, emergency responders should screen people for external contamination and decontaminate them in accordance with their emergency decontamination procedures. Following external decontamination, emergency responders should screen people for internal contamination as described later in this Report (Section 8). Realistically, however, people who are not critically injured may self-evacuate if they have their own transportation. To prevent hospitals from being overwhelmed by people who seek treatment or screening for contamination, people who self-evacuate should be directed to reception centers (CDC, 2007) for screening and decontamination, or they should be directed to go home to shower using plain soap and water and to report later to a reception center to be screened for internal contamination. Depending on the size of the incident (i.e., the number of people involved), screening may have to be spread over several days. As emergency responders approach the scene of the incident to initiate their response, they should conduct a radiological assessment of the scene and protect themselves as described below. 6.1 General Guidance for Emergency Responders NCRP Commentary No. 19 (NCRP, 2005) and NCRP Report No. 161 (NCRP, 2008a) provide emergency responders with guidance on how to conduct their work in a potentially-dangerous radiation environment. This Report recommends that emergency 53

54 / 6. RADIOLOGICAL TRIAGE AND SCREENING GUIDANCE responders follow the guidance provided in these previous publications, and it provides a summary here for convenience. Emergency responders should have a basic understanding of external radiation exposure and external and internal radioactive contamination. They should know what contamination is and the processes necessary to mitigate contamination to alleviate potential medical consequences. They should also know the appropriate measures to protect themselves, others onsite, and members of the public. Once it is known that the incident involves radionuclide dispersion, emergency responders should be assigned radiation-survey equipment to record exposure rates and the cumulative dose they are receiving. The equipment specifications are discussed in NCRP Report No. 161 (NCRP, 2008a). The main protective action will be medical treatment for those injured, which may include evacuation from the affected area, decontamination of those who were contaminated, transport, and assessment of external and internal exposures. Emergency responders entering the area should be prepared to work in a contaminated environment, and those leaving the area should be screened and decontaminated as necessary. The incident site should be segmented into two radiation-control zones: hot zone and dangerous radiation zone (NCRP, 2010). An outer perimeter separates the uncontrolled zone from the hot zone and is established at an exposure rate of 0.1 mGy h–1 (~10 mR h–1), 1,000 Bq cm–2 (60,000 dpm cm–2) for beta and gamma surface contamination, or 100 Bq cm–2 (6,000 dpm cm–2) for alpha surface contamination. An inner perimeter separates the hot zone from the dangerous radiation zone and is established at an exposure rate of 0.1 Gy h–1 (~10 R h–1). Within the outer perimeter emergency responders should evacuate members of the public, isolate the area, and minimize time spent in the area. Within the inner perimeter emergency responders should limit their actions to time-sensitive, mission-critical activities such as lifesaving. NCRP Report No. 161 (NCRP, 2008a) provides additional important guidance on protection of emergency responders. 6.1.1

Selecting an Appropriate Radiation Survey Instrument

Radiation and contamination surveys should be performed using a radiation detector that is sensitive to the type of radiation(s) present and is attached to a meter that reads out in appropriate units. For example, a sodium iodide detector is designed to detect photon radiation and should not be used to survey for pure beta-emitting radionuclides such as 90Sr or pure alpha-emitting radionuclides. Results of surveys for external surface contamination should be reported in units of becquerel or counts per minute

6.1 GENERAL GUIDANCE FOR EMERGENCY RESPONDERS

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(cpm), and surveys of external radiation exposure rates should be reported in units of mGy h–1 or mR h–1. Thus, to survey for 137Cs surface contamination, one might use a thin-window GM pancake probe attached to a meter that reads in counts per minute, while surveying for radiation exposure-rate levels would call for using an ion chamber, microR meter, or sodiumiodide detector that is attached to a meter reading in mGy h–1 or mR h–1. Before the record is filed and based on the detection efficiency of the detector, any readings in counts per minute should be converted to becquerel or disintegrations per minute (dpm) or Bq cm–2 or dpm cm–2 for future dose reconstruction. Further information on meter and detector selection criteria is provided in Section 8, especially in Table 8.1. In general, people will be surveyed for the presence of external radioactive contamination and the results should be noted in units of counts per minute. These readings will be compared to screening criteria using the flowcharts described later in this section to determine how to proceed: (1) at the scene of an incident, and (2) at the hospital, respectively. The Conference of Radiation Control Program Directors, Inc. (CRCPD) developed a guide for first responders that contains many procedures for responding to an RDD, including one for monitoring people for contamination (CRCPD, 2006). Similarly, CDC published a population monitoring guide for local and state public-health planners that describes procedures for monitoring a population after a radiological incident (CDC, 2007). 6.1.2

Presurvey Radiation Survey Instrument Checks

Several quick checks are required before using any radiation survey instrument to ensure that the instrument is operational. These checks are: • Confirm that the detector and meter are appropriate for the survey to be performed. • Confirm that the meter has been calibrated for the radionuclide(s) of interest within the last year by checking the calibration sticker on the instrument. If the instrument has not been calibrated, it may be used in an emergency, but such use should be noted in the screening records. • Quickly inspect the instrument and cable to confirm that they are in useable condition and properly attached. • Confirm that the batteries are charged by pressing the battery test button (or by setting the instrument switch to the battery test position) and noting that the needle deflects to the appropriate position.

56 / 6. RADIOLOGICAL TRIAGE AND SCREENING GUIDANCE • Zero an ion-chamber survey meter if it has a zero setting. • Turn the instrument to the lowest switch position that gives a usable reading [that is a reading between about one-fifth and four-fifths of the scale (e.g., between one and four on a scale that goes from zero to five)]. Multiply the needle reading by the switch position (e.g., if the needle reads 230 and the switch is on the “× 100” scale then the meter reading is 23,000). This reading is the background radiation level or count rate. All of the measurements will be compared to current background radiation levels. Record the background readings for future reference. • Use a check source or other radiation source to confirm that the instrument detects radiation. 6.1.3

Surveying for Radioactive Contamination

• Perform pre-use checks, including recording the background count rate. • Hold the detector ~1 cm (0.5 inch) from the person to be surveyed. • When surveying people, spot-check the mouth and nose, chest, hands and fingers, knees, and feet; and scan other areas if time permits. • To survey a specific location (e.g., the nose or fingertips), hold the survey meter over the spot to be surveyed for at least 10 s. • To survey large areas, move the detector at a slow speed [~3 to 5 cm s–1 (1 to 2 inches s–1)]. • Note the highest count rate (in counts per minute above background levels) and record. 6.2 Radiological Triage and Screening Procedures Before potentially-contaminated people are transported or allowed to enter the emergency department, they should receive some degree of radiological screening, if personnel and equipment resources permit. When the survey shows that they are externally contaminated, they should be decontaminated in accordance with local protocol. However, critically-injured persons should be treated without regard to radioactive contamination and transported directly to the hospital after onsite stabilization. It may also be advisable to establish triage screening stations at hospitals for those who are injured and self-refer or self-evacuate. Finally, those who are uninjured and who self-evacuate, and those who are not at the scene of the incident but who are downwind or have reason to

6.3 INITIAL SCREENING AT SCENE

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believe they are contaminated may require screening after the emergency is over. For this purpose there may be a need for largescale screening at a suitable location other than the hospital (e.g., sports arena). Radiological triage and screening should be conducted by properly trained personnel using appropriate radiation survey instruments. Training need not be extensive; it is possible to learn how to perform a screening survey in a short time following the procedures provided here and elsewhere in this Report. People who test positive for external or internal contamination should be asked whether they have had a recent nuclear medicine exam or have received radioactive implants such as 125I seeds. Careful screening of these patients may be necessary to determine whether high survey meter readings are due to contamination or medical radionuclides. 6.3 Initial Screening at Scene Triage and initial screening at the scene of a radiological or nuclear incident should be limited to surveys that can be performed without placing people at risk from their injuries. Patients who have suffered life-threatening injuries should be given medical care immediately, without regard to contamination. Caring for radionuclide-contaminated people will almost never put caregivers at risk provided they are using techniques equivalent to standard precautions (infection prevention practices such as use of gloves, gown, mask, eye protection, which apply to all patients, regardless of suspected or confirmed infection status). It is highly unlikely that the radioactive materials from a contaminated patient, or the radiation emitted by this radioactive material will place a medical caregiver at risk. In fact, with the exception of a patient containing embedded fragments of radioactive material (e.g., shrapnel), medical caregivers who take standard precautions are unlikely to receive a significant dose of radiation (Smith et al., 2005). As noted in the Rojas-Palma et al. (2009) handbook, radiological triage procedures to be followed depend on the circumstances of the exposure (e.g., whether significant external exposure may have occurred due to direct exposure or whether most of the exposure may be due to environmental or personal contamination). A general flowchart for screening at the scene of a large incident such as an RDD incident is shown in Figure 6.1. Those affected individuals

58 / 6. RADIOLOGICAL TRIAGE AND SCREENING GUIDANCE

Fig. 6.1. Flowchart for screening potentially-contaminated persons at the scene of a large-scale radiological or nuclear incident.

whose medical conditions are not life-threatening should receive contamination screening to the extent that their medical conditions, the availability of survey equipment, weather conditions, and personnel resources permit. For example, screening thousands of people may take days, requiring advance planning, logistical support, and communication and follow-up activities. In some cases, exposure to the weather may prove a higher acute health risk than the potential radiological hazards, if the initial screening should be conducted outdoors. The recommended screening value when surveying with a “pancake-type” GM detector is 1,000 cpm. Persons with more than this level of contamination on their skin should be decontaminated when it is possible and appropriate to do so, and persons with less than this level of contamination may be sent home (if appropriate) or treated without regard to their contamination. This level of contamination is appropriate for intact skin, wounds, and bodily orifices such as the mouth and nose. The justification for this value is discussed below. The area of a typical pancake-type detector is 15 cm2. If the counting efficiency of the detector is 10 %, then 1,000 cpm in the area beneath the detector corresponds with a total of 10,000 dpm over 15 cm2 (~670 dpm cm–2). Calculations performed using VARSKIN 3

6.3 INITIAL SCREENING AT SCENE

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(Durham, 2006) indicate that skin dose from 90Sr (a high-energy beta emitter) contamination is 2.5 × 10–2 μSv h–1 (2.5 × 10–3 mrem h–1) for 1 dpm cm–2, so 670 dpm cm–2 result in a skin dose of ~16.5 μSv h–1 (1.65 mrem h–1). This level of exposure will produce a radiation dose of <0.5 mSv (50 mrem) to the skin even if left unwashed for 24 h and poses no physical risk to the contaminated person. Assuming more realistic counting efficiencies for high-energy beta emitters (many pancake-type GM probes are 30 to 40 % efficient for this energy of beta radiation) reduces the dose further. An action level of 1,000 cpm should remain protective of health from uptake through broken skin (e.g., wounds) or through the mouth or nose. The amount of radioactivity contained in 670 dpm is ~112 Bq (~4 nCi). This level of uptake is much lower than the most limiting CDG listed in Table 3.14 of NCRP Report No. 161 [5,200 Bq (0.14 μCi) for 227Ac] (NCRP, 2008a). In addition to the above, it may not be possible to conduct comprehensive screening of ambulatory persons at or near the scene of a radiological incident because many of those able to walk may selfevacuate. This may have a significant immediate impact on healthcare organizations, and some who self-evacuate may refer themselves for medical care or for radiological screening after the emergency phase has ended. These situations are both described below. Immediately after a radiological or nuclear incident, members of the public who are not injured will most likely self-evacuate. Many of these people may be worried about possible radioactive contamination and will likely seek help at the nearest hospital even if they were not directly involved in the incident. To prevent hospitals from becoming overwhelmed with worried but uninjured members of the public, CDC recommends that communities establish reception centers to screen people for contamination, to decontaminate them, and to enroll them in a registry for long-term health follow-up (CDC, 2007). These centers could be setup in community-wide locations such as fair grounds, sports facilities, or high schools. To be effective, these reception centers need to be organized to allow them to be activated quickly, and timely public announcements would be necessary to direct the public to these reception centers. Areas with high levels of radiation should be cordoned off to minimize radiation exposure to those awaiting screening as well as those conducting this screening. To reduce interference from radioactive contamination at the scene of the incident, it may be advisable to screen people in an unaffected area, at the greatest practical distance from the site of the incident. To that end, the incident

60 / 6. RADIOLOGICAL TRIAGE AND SCREENING GUIDANCE commander may consider identifying locations where hundreds or thousands of potentially-contaminated persons may be screened and developing ways to transport large numbers of people to these locations. Alternately, the incident commander may simply decide to permit many uninjured people to leave the scene of the incident and to shower at home, accepting that there may be a spread of contamination through the city (Appendix E). In such cases, radiological screening will only be required for those people requiring evacuation to medical care facilities, unless the urgency of their medical condition precludes screening. Those making the decision to attempt to control the spread of contamination by establishing boundaries and controlling the movements of contaminated persons should be aware that this decision may prove controversial and difficult to enforce. As noted above, many who are uninjured may self-evacuate before boundaries are established. In addition, attempting to restrict the movements of large numbers of people may lead to unrest and anger. Maintaining these boundaries may be difficult, even with the assistance of law-enforcement personnel. 6.4 Initial Screening at Hospital It is likely that many potentially-contaminated individuals will reach the hospital by foot, in private vehicles, or by other means that do not rely on emergency response vehicles. Also, they may arrive at the hospital prior to the emergency vehicles. As noted above, the critically injured and persons with life-threatening medical conditions should receive immediate care without regard to their contamination. All others should be triaged, screened and decontaminated if personnel, equipment, space, and the person’s medical condition permit (Figure 6.2). People who were not at the scene of the incident but may be worried that they are contaminated should be directed to a secure secondary location for screening and reassurance. The first priority should be to stabilize those whose injuries are life-threatening; followed by external decontamination; and depending on the magnitude of the incident, evaluation for radionuclide intake, decorporation (if necessary) of any internal radionuclides and screened for the acute radiation syndrome (ARS). External contamination may interfere with determining the presence of internal activity. For this reason, external decontamination should be completed before more sensitive surveys such as chest counting are undertaken.

6.6 BIODOSIMETRY

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6.5 Mass Screening Following the Emergency Phase In the aftermath of a radiological or nuclear incident, as news of the nature of the attack becomes known, many citizens may be concerned about the potential for contamination. These persons may wish to be surveyed for contamination. In the case of the contamination incident in Goiânia, Brazil, over 110,000 citizens voluntarily reported to the local soccer stadium for radiological screening when the extent of the contamination became known (IAEA, 1998). It is reasonable to assume that any community is likely to experience similar levels of concern in the aftermath of a radiological or nuclear incident. 6.6 Biodosimetry This Report is intended primarily to provide advice on the detection and management of persons who have internal radionuclide depositions resulting from an RDD or IND incident. The relative importance of radiation exposure from an internal deposition of one or more radionuclides compared with that received from an external source differs considerably for RDD and IND incidents. In an RDD incident, the primary means by which members of the population may be exposed to radiation will be from internally-deposited radionuclides except for a small fraction that may have been exposed externally to radiation from fallout very close to the detonation site of an RDD. On the other hand, an IND produces large amounts of external radiation exposure delivered at a high dose rate over a short period of time, both during the nuclear detonation and later from fallout near the point of detonation. Radionuclide intakes can occur later from the fallout downwind and are of less overall importance related to early occurring deterministic effects of radiation. It is important that physicians be aware of these differences and the relative importance of brief radiation exposure from an external source and prolonged internal irradiation of different body organs. Much of the information in this Report is directed to dealing with the latter case for internally-deposited radionuclides. In this case, in vivo and in vitro bioassay procedures are the best available tools for estimating radionuclide intakes over a broad spectrum of radionuclides in different physical and chemical forms. These intake values can be compared with appropriate CDG intake values as one means of determining whether further medical intervention should be considered. In the case of an IND incident, it will be known that some people have been exposed to whole-body irradiation from an external

62 / 6. RADIOLOGICAL TRIAGE AND SCREENING GUIDANCE

Fig. 6.2. Flowchart for screening potentially-contaminated persons at the hospital (adapted from NCRP, 2008a).

6.6 BIODOSIMETRY

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64 / 6. RADIOLOGICAL TRIAGE AND SCREENING GUIDANCE source. It is crucial that individuals who have received absorbed doses approaching the lethal dose for causing death in 50 % of exposed persons (LD50) be identified as soon as possible after exposure because medical intervention (administration of antibiotics and platelet and cytokine treatments) roughly double the LD50 (Anno et al., 2003). In addition, for RDD incidents, physicians should consider conducting biodosimetry on some patients, particularly those with high levels of external radionuclide contamination and those who were close enough to the device to have been exposed to a potentially high dose of external radiation. Physicians may find use of a biodosimetry worksheet helpful (AFRRI, 2007). Besides screening individuals for internal contamination, total effective dose can be estimated with other methods including analysis of dicentric chromosomes and chromosome translocations, electron paramagnetic resonance with teeth as well as other methods (ICRU, 2002).

7. Clinical Decision Guide: Concept and Use 7.1 Clinical Decision Guide Concept One of the challenging medical management aspects of treating persons having radionuclide intakes is to determine at what level of intake dose intervention therapy, as described in Section 9, should be considered. In NCRP Report No. 161 (NCRP, 2008a) a new operational guide, the Clinical Decision Guide (CDG), was defined to help the physician in this process. The CDG provides an important measure that physicians should use when considering the need for medical treatment of individuals having an internal radionuclide deposition. The CDG is the intake level for a radionuclide that, if exceeded, may justify medical treatment to decorporate (i.e., enhance excretion of) the radionuclide. Early reduction of the body content of a radionuclide will reduce the total dose that will be received from this intake and thereby reduce or possibly prevent the occurrence of early effects of radiation (e.g., tissue damage due to cell death), and will reduce the likelihood of late effects (e.g., cancer). The CDG for intake of a given radionuclide has been determined using calculations based on the physical and chemical form of the radionuclide, the route of exposure, and the subsequent disposition of the radionuclide within the body and its excretion from the body. Numerical values of dose used as the bases of the computed CDG intake values for different radionuclides in this Report were established using recommendations and limits for emergency situations and current knowledge of dose thresholds for early occurring tissue damage (known as deterministic effects) and the risks of late-occurring cancers (known as stochastic effects). For all radionuclides, other than the isotopes of iodine, the doses in adults are 0.25 Sv (25 rem) (50 y committed effective dose) for consideration of late-occurring cancers; a 30 d RBE-weighted absorbed dose value of 0.25 Gy-Eq (25 rad-Eq) for consideration of bone marrow 65

66 / 7. CLINICAL DECISION GUIDE: CONCEPT AND USE damage; and a 30 d RBE-weighted absorbed dose value of 1 Gy-Eq (100 rad-Eq) for consideration of pulmonary damage.2 For intake of a radionuclide other than iodine radionuclides, the CDG for an adult is the maximum intake satisfying all three of these dose constraints based on tissue damage and cancer. Table 7.1 lists the adult CDG values for radionuclides covered in this Report (NCRP, 2008a). More CDG information on these radionuclides is given in the CDG fact sheets in Sections 7.3 to 7.12 and in Section 7.13. 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, reflecting the increased vulnerabilities during development and maturation (AAP, 2003). Children weighing >70 kg should be considered as adults. The CDG values for an intake or expected intake of radioiodine are determined on a different basis, the limitation of dose to the thyroid following guidance given by FDA (2001) for administration of KI to persons of different ages and conditions. Thyroid blockade by administration of KI can substantially reduce thyroid uptake of radioactive iodine if administered 1 to 2 d before exposure or within 4 to 5 h after exposure. Information on the age and gender basis of CDG values for 131I are given in Sections 7.6 and 7.13. The CDG is 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 estimates of the activity taken into the body following exposure to radionuclides often involve significant errors, as revealed by more detailed follow-up measurements. For radiation dose estimates of the magnitude indicated in the definition of the CDG, the increase in risk of stochastic effects and conceivably even early deterministic effects may become an important consideration with regard to a clinical course of action. Clinical actions based on the CDG will vary, however, depending on available time and resources (as determined, for example, by the 2RBE

(relative biological effectiveness) is a factor used to compare the biological effectiveness of absorbed doses from different types of ionizing radiation, determined experimentally. RBE is the ratio of the absorbed dose of a reference radiation to the absorbed dose of the radiation in question required to produce an identical biological effect in a particular experimental organism or tissue. In these dose calculations for deterministic effects, RBE values of two and seven were used for alpha-particle radiation in the bone marrow and lungs, respectively (Scott, 1993).

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7.1 CLINICAL DECISION GUIDE CONCEPT

TABLE 7.1—Adult CDG values for radionuclides considered in this Report (NCRP, 2008a).a Radionuclide 60

Co

90

Sr

131 c

I

137

Cs

192

Ir

226

Ra

238

U

d

238

Pu

239

Pu

241

Am

Exposure Route

Typeb

MBq

μCi

Inhalation

M

35

950

Inhalation

S

15

400

Inhalation

F

8.3

230

Ingestion

Soluble

8.9

240

Inhalation

Vapor

0.26

6.9

Ingestion

Soluble

0.26

6.9

Inhalation

F

58

1,600

Ingestion

Soluble

28

760

Inhalation

M

59

1,600

Inhalation

S

50

1,400

Inhalation

M

0.11

3.1

Inhalation

M

0.12

3.2

Inhalation

S

0.037

0.99

Inhalation

M

0.0081

0.22

Inhalation

S

0.023

0.61

Inhalation

M

0.0076

0.20

Inhalation

S

0.030

0.80

Inhalation

M

0.0093

0.25

a CDG values for other radionuclides are available in NCRP Report No. 161 (NCRP, 2008a). bTypes F = fast, M = moderate, and S = slow rate of absorption from the lung (ICRP, 1994a). c For adults 18 to 40 y of age. See Section 7.6 for CDG values for other age ranges. d Table entries for 238U may also be applied to 234U and 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 (Section 7.2.5).

68 / 7. CLINICAL DECISION GUIDE: CONCEPT AND USE number of exposed persons and the extent of life-threatening injuries). For example, 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 tissue-specific absorbed doses over different time periods. The CDG is not intended to instruct physicians on a specific course of action. Rather, CDGs are intended as a tool to be used to help a physician determine when a radionuclide intake may have clinical significance.

It seems reasonable to assume that for an exposure of <1 CDG, the attending physician may conclude that the radiological risk to the patient does not warrant decorporation therapy, especially if there are large numbers of injured patients to attend to. However, it is not unreasonable to assume that if the number of other patients is low, a physician may decide to commence decorporation therapy for an intake of <1 CDG. In fact, this latter course of action is typical of industrial incidents in which there are only a few contaminated persons. Also, a physician may decide to begin decorporation therapy without knowing whether the CDG has been exceeded. For example, a patient with extensive injuries and external contamination who was located near the center of the radiological or nuclear incident may be assumed to be internally contaminated at a significant level and likely would benefit from early decorporation therapy. 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, the biochemical and physical properties of the internally-deposited radioactive material, and its site(s) 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. 7.2 Clinical Use of the Clinical Decision Guide 7.2.1

Decision-Making Process

For the purposes of this Report, it is assumed that the attending physician is confronted with a large number of patients who may or may not have experienced an intake of radioactive materials. In such an instance, it is essential that the physician be able to quickly and accurately determine which patients are at short- or long-term

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risk from internally-deposited radioactive material. Figure 7.1 shows a process for making decisions on whether to begin decorporation therapy. Section 9 describes the use of the CDG in helping to determine a course of treatment for the patient. The CDG is defined for adult patients. Children and women who are pregnant or nursing should be limited to one-fifth (20 %) of a CDG, and the values presented in this section should be modified accordingly for these patients. Thus, women who are pregnant or nursing children may be expected to receive decorporation therapy at a lower level of internal contamination than other adults (NCRP, 2008a). 7.2.2

Use of the CDG Tables

Chest or whole-body counts or measurements of activity in excreta or on nasal swabs may be used to estimate the level of intake to which a patient has been exposed. The fact sheets in Sections 7.3 to 7.12 give levels of measured activity in the 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 exposure listed. The CDG values in each fact sheet were drawn from Tables 11.1 and 11.2 in NCRP Report No. 161 (NCRP, 2008a).

Fig. 7.1. Example of a screening process to determine whether a patient equals or exceeds the CDG and the use of decorporation therapy should be considered.

70 / 7. CLINICAL DECISION GUIDE: CONCEPT AND USE 7.2.3

Use of a Single-Void Urine Sample Collected During the First 24 h

It may only be possible to collect a single-void sample during the first 24 h after an RDD or IND incident. However, the tabulated values given in the CDG fact sheets that follow are based on 24 h urine collections. The following items provide guidance on how these spot samples should be collected and used. 1. Urine samples collected within 2 h of a radionuclide intake may not be representative of the intake (possible dilution by urine in the bladder prior to the exposure). For accurate assays, the patient should void and then collect a sample over the following 2 to 4 h. If early detection of a significant intake is important, an earlier sample may be used with the understanding that it may significantly under-represent the intake. 2. Four different methods for extrapolating data from a spot urine sample to a 24 h value are given in NCRP (2008a). They are based on four characteristics of the spot sample: - volume of urine; - collection time interval relative to a 24 h sample; - creatinine content; and - specific gravity. Items 3 and 4 both require additional measurements on the sample. The simplest and quickest correction to make in a mass casualty situation is based on the volume of urine collected in the sample and its relationship to the 24 h reference value for the gender and age of the subject as given in Table 7.2. 3. As an example of using Method 1, assume that an adult female subject provided a spot 80 mL sample and the radionuclide content was estimated to be 0.037 MBq (1 μCi). The content of a 24 h sample based on this information and the data in Table 7.2 would be [0.037 MBq (1 μCi) × 1,200 mL] ÷ 80 mL = 0.56 MBq (15 μCi) or 3.3 × 107 dpm. 4. As an example of Method 2, assume that a subject provided a sample collected over a 5 h period that had a measured content of 0.048 MBq (1.3 μCi). The estimated content of a 24 h sample would be 0.048 MBq (1.3 μCi) × 24 ÷ 5 = 0.23 MBq (6.2 μCi) or 1.4 × 107 dpm. 7.2.4

Using the CDG with an Intake of Multiple Radionuclides

It is possible that some persons may be exposed to more than one radionuclide. In such cases, it is appropriate to apply the “sum

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TABLE 7.2—Reference values for 24 h urine volume (ICRP, 2002b). Excretion (mL d–1) Age Male

Newborn

Female

300

1y

400

400

5y

500

500

10 y

700

700

15 y

1,200

1,200

Adult

1,600

1,200

of fractions” method to determine the effective intake. Say, for example, that a person has experienced an intake of three radionuclides: A, B and C, as shown in Table 7.3 and discussed in greater detail in Section 11.1. The total intake, using the sum of fractions method would be 1.3 CDG, and the attending physician would be advised to consider beginning decorporation therapy once other medical conditions had been stabilized. Information of CDG values for radionuclides beyond those considered in this Report is available in NCRP Report No. 161 (NCRP, 2008a). 7.2.5

Determining an Intake for Times More than 24 h in the Past

The tables of CDG values presented in this Report and in NCRP Report No. 161 (NCRP, 2008a) assume that the intake occurred within 24 h of the time that the assessment is made. While this is appropriate for patients who arrive at an emergency department and are evaluated quickly, it is possible that more than 1 d may pass before an evaluation can be made. In the case of an RDD, for example, several days or weeks may pass between the time of the incident and the time that it is discovered. It is also reasonable to assume that many affected individuals may self-evacuate from the scene before discovering the presence of radioactive materials; such persons may return for evaluation in the days or weeks following the incident. For example, Table 7.4 for 241Am has been reproduced from Table 20.7 of NCRP Report No. 161 (NCRP, 2008a). The retention and excretion fractions used to calculate the 24 h values from which the CDG can be determined were taken from this table. The inhalation CDG for 241Am is 9.3 kBq (0.25 μCi) (NCRP, 2008a).

72 / 7. CLINICAL DECISION GUIDE: CONCEPT AND USE TABLE 7.3—Example of a CDG calculation involving more than one radionuclide (Section 11.1). Radionuclide

CDG (MBq)

A

1

B C

Intake (MBq)

Fraction of CDG

0.2

0.2

10

5

0.5

20

12

0.6

Sum

1.3

TABLE 7.4—Reference values for retention and excretion of 241Am (percentage of intake) as a function of time after acute inhalation of moderately soluble form (Type M) by an adult; particle size = 5 μm (AMAD).a,b Day After Intake

24 h Urinary Excretion (%)

24 h Fecal Excretion (%)

Retained in Lungs (%)

Retained in Body (%)

1

1.8 × 10–1

1.1 × 101

5.8 × 100

5.0 × 101

2

2.3 × 10–2

1.5 × 101

5.6 × 100

2.6 × 101

3

1.3 × 10–2

8.0 × 100

5.5 × 100

1.5 × 101

5

7.2 × 10–3

1.3 × 100

5.3 × 100

9.1 × 100

7

5.8 × 10–3

2.3 × 10–1

5.2 × 100

8.2 × 100

10

4.9 × 10–3

5.7 × 10–2

5.0 × 100

7.9 × 100

15

3.9 × 10–3

4.2 × 10–2

4.6 × 100

7.6 × 100

20

3.3 × 10–3

3.7 × 10–2

4.3 × 100

7.4 × 100

30

2.6 × 10–3

2.8 × 10–2

3.8 × 100

7.1 × 100

40

2.3 × 10–3

2.1 × 10–2

3.4 × 100

6.8 × 100

50

2.0 × 10–3

1.7 × 10–2

3.1 × 100

6.6 × 100

60

1.9 × 10–3

1.3 × 10–2

2.8 × 100

6.4 × 100

70

1.8 × 10–3

1.0 × 10–2

2.6 × 100

6.3 × 100

80

1.7 × 10–3

8.2 × 10–3

2.4 × 100

6.2 × 100

90

1.6 × 10–3

6.6 × 10–3

2.2 × 100

6.1 × 100

100

1.5 × 10–3

5.4 × 10–3

2.0 × 100

6.0 × 100

aThis table was calculated for an exposed adult worker in NCRP Report No. 161 (NCRP, 2008a) but the numbers also apply to an exposed adult member of the public that inhaled particles having the same characteristics. bAMAD = activity median aerodynamic diameter of a log-normal size distribution of airborne particles.

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The second column of Table 7.4 shows us that 0.18 % of inhaled Am (solubility Type M) will be excreted in the urine in the first 24 h post-exposure. Using these values, it can be determined that a 24 h urine sample collected for the first 24 h post-exposure will contain 9.3 kBq × 1.8 × 10–3 = 1.7 × 10–2 kBq (0.25 μCi × 1.8 × 10–3 = 4.5 × 10–4 μCi) of 241Am if the patient had an intake by inhalation of 1 CDG of 241Am. However, if the patient arrived at the hospital for evaluation one week after exposure, some of the 241Am will have already cleared the body, and much of what remains will be in the lung, liver and skeleton. Excretion of the 241Am through the urine slows greatly after the first few days post-exposure. Accordingly, it is necessary to calculate the intake corresponding to an intake of 1 CDG, even when the bioassay takes place some time after exposure. In Table 7.4, we see that at 7 d after an acute inhalation exposure to 241Am in a moderately soluble form, a 24 h urine sample will contain 5.8 × 10–3 % of the radioactive material that was inhaled. Thus, a 24 h urine sample would be expected to contain 9.3 kBq × 5.8 × 10–5 = 5.4 × 10–4 kBq (0.25 μCi × 5.8 × 10–5 = 1.5 × 10–5 μCi). Similar calculations can be made for other radionuclides at other times after intake. 241

7.2.6

Special Considerations for Uranium CDGs

Table 11.1 of NCRP Report No. 161 (NCRP, 2008a) gives CDG values for a broad range of radionuclides, including those covered in this Report, based on their radiological properties. A footnote to the CDG values for 238U inhaled in absorption Type M or S aerosols states, “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 (DU) but not necessarily for inhalation of enriched uranium.” Thus, for the exposure of a large group of people as considered in this Report, it is important to understand when radiological toxicity may be the limiting case and when chemical toxicity may be the limiting case (Fisher et al., 1990; Just, 1984; Just and Emler, 1984; Morrow et al., 1982). There has been renewed interest in the behavior of inhaled DU aerosols in this country since several U.S. fighting vehicles were hit by friendly-fire munitions containing DU during the Gulf War in 1991. Soldiers in, or near the vehicles were exposed to airborne DU and some still carry embedded fragments of DU in their tissues. Research since that time has been focused on evaluations of the chemical and radiological risks from internal depositions of DU by inhalation and wound exposures as described in publications

74 / 7. CLINICAL DECISION GUIDE: CONCEPT AND USE such as Guilmette and Cheng (2009), Guilmette et al. (2009), Hahn et al. (2009), Kathren and Burklin (2008), Leggett (2006), Leggett and Pellmar (2003), McDiarmid et al. (2006), Parkhurst et al. (2009), Roszell et al. (2009), Squibb and McDiarmid (2006), and Squibb et al. (2005). Leggett (2006) provided a detailed review of the biokinetics of uranium in the human body and, using current ICRP models, modeled the absorption of uranium from the lung resulting from a chronic daily inhalation exposure to uranium. These analyses provided a comparison of chemically- and radiologically-based limits on the intake for chronic inhalation of soluble (Type F), moderately soluble (Type M), and relatively-insoluble (Type S) forms of uranium containing different levels of 235U (i.e., depleted, natural or enriched forms). Leggett later used a similar approach to construct the curves given in Figure 7.2 for a population receiving an acute inhalation exposure to uranium such as might occur in an RDD or IND incident. This figure is based on an acute inhalation exposure scenario involving an aerosol having a log-normal particle-size distribution with an activity median aerodynamic diameter (AMAD) of 5 μm and absorption Types F, M or S. The three lines marked “Chem” are limits on intake based on a chemical nephrotoxic limit of 3 μg g–1 kidney and the three lines marked “Rad” are CDG intake values for an adult resulting in a 50 y committed effective dose of 0.25 Sv based on the radiological properties of uranium. For a given absorption Type X, the limiting factor (radiological or chemical risk) is determined by the lower of the curves labeled “Chem X” and “Rad X” at the appropriate 235U percentage by weight. These graphs show where chemical nephrotoxicity is limiting and where stochastic risk from radiation dose is limiting. For these cases, chemical nephrotoxicity is always limiting for Type F (5 μm), and radiological toxicity is always limiting for Type S (5 μm). For Type M (5 μm), chemical nephrotoxicity is limiting up to ~30 % 235U enrichment and radiological toxicity is limiting for higher enrichment. Although this graph shows curves for radiological CDGs for inhaled uranium, specific CDG values have not been established for chemical toxicity as a guide for physicians considering various treatment options. However, information from the Capstone studies discussed below give some general guidance on this topic. The Capstone series of experiments (Parkhurst and Guilmette, 2009; Parkhurst et al., 2004) studied the impacts of various aerosol properties on health risk assessments for inhaled DU aerosols produced at high temperatures.3These studies were conducted to provide DU aerosol exposure ranges to personnel in armored combat vehicles perforated by DU munitions and to apply these data in

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Fig. 7.2. Limiting intake of uranium based on a chemical (Chem) nephrotoxic limit of 3 μg g–1 and a radiological (Rad) CDG value of 0.25 Sv committed effective dose (50 y) for an acute inhalation exposure to uranium (5 μm AMAD). Curves are based on the current ICRP model for uranium in adults. The acute intakes giving the limiting kidney concentration (chemical limit) are 0.029 g for Type F, 0.23 g for Type M, and 7.4 g for Type S aerosols.3 The specific activity of uranium increases with enrichment, not because of the replacement of some 238U [radioactive half-life (T1/2) = 4.5 × 109 y)] with 235U (T1/2 = 7.0 × 108 y), but primarily because of the increase in the amount of 234U present (T1/2 = 2.5 × 105 y) (Hinnefeld et al., 1988).

3Leggett,

R.W. (2010). Personal communication (Oak Ridge National Laboratory, Oak Ridge, Tennessee).

76 / 7. CLINICAL DECISION GUIDE: CONCEPT AND USE assessments of human health risks to these exposed personnel. Aerosol measurements included airborne concentrations, particlesize distributions, and characterizations of physicochemical properties including in vivo solubility estimations based on dissolution in simulated lung fluids. The tests related to personnel in or on Abrams tanks or Bradley fighting vehicles (Parkhurst and Guilmette, 2009; Parkhurst et al., 2004). Results of these studies are divided into two major divisions: Part I, Depleted Uranium Aerosol Characterization; and Part II, Human Health Risk Assessment. These results from tests in which DU munitions were fired through DU and non-DU armor on these vehicles provide important information that can be used when considering the relationships between aerosol characteristics and human health risk issues in an RDD or IND incident. The in vivo absorption types of these aerosols were usually in the Type M and sometimes Type S categories (Guilmette and Cheng, 2009). One of the factors influencing these results was the type of armor being impacted by the DU penetrator. A peak uranium concentration of 3 μg g–1 kidney required an intake of ~250 mg uranium from penetration of conventional vehicle armor and ~650 mg for penetrated DU armor (Guilmette and Cheng, 2009; Parkhurst et al., 2004). The lower end of this range of intakes agrees quite well with that obtained by Leggett in Figure 7.2 for the uranium intake required to produce a peak uranium concentration of 3 μg g–1 uranium in kidney based on Type M absorption. In Part II of the Capstone study, Roszell et al. (2009) assessed the toxicity of Capstone DU oxides and other uranium compounds. With regard to a previous health risk assessment related to DU aerosols in the Gulf War, they stated that conclusions in an earlier chemical toxicity assessment were limited to whether or not a guideline was exceeded; there were no conclusions as to whether there may have been renal damage if a guideline was exceeded. To address this latter point, the authors developed a risk model equation that permitted them to use uranium intake and health effects data to model peak uranium concentrations in the kidney and resulting health effects from human exposure cases in the literature. They used their models to establish four renal effects groups having different levels of severity and outcomes as shown in Table 7.5. A renal toxicity concentration of 3 μg g–1 falls in the renal effects group 1 level as does a concentration twice that, 6 μg g–1. Although no CDG values have been established related to the chemical nephrotoxicity of inhaled uranium, the information provided in Figure 7.2 and Table 7.5 are useful for a physician considering the need for various treatment options. In Figure 7.2, the

Possible protracted indicators of renal dysfunction Possible severe clinical symptoms of renal dysfunction

≥6.4 to ≤18 ≥18

2

3

REG = renal effects group, a scale of renal toxicity.

a

Possible transient indicators of renal dysfunction

≥2.2 to ≤6.4

1

No detectable effects

≤2.2

0

Acute Renal Effect

Kidney Uranium Concentration (μg g–1 kidney tissue)

REGa

Likely to become ill

May become ill

Not likely to become ill

No clinically detectable effects

Predicted Outcome

TABLE 7.5—Characterization of the chemical risk for uranium in the kidney (Roszell et al., 2009).

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78 / 7. CLINICAL DECISION GUIDE: CONCEPT AND USE intersection of the line for Chem M based on current ICRP models with the y axis shows that a renal concentration of 3 μg g–1 uranium in kidney results from an intake of ~230 mg of uranium. Other levels of uranium intake would scale in the same proportion. For instance, doubling the uranium intake value to 460 mg would also double the resulting peak uranium concentration in the kidney. Therefore, initial estimates of the percent 235U in the exposure aerosol, intake of uranium in milligrams, and absorption type (F, M, S), the physician can calculate an estimated peak concentration of uranium in the kidney. This number can then be used with the information in Table 7.5 to consider various medical treatment options. Section 20.24 of NCRP (2008a) provides additional important information on the properties of uranium radionuclides, their biokinetics, and internal dosimetry that relate to this topic. 7.3 Americium-241: Clinical Decision Guide Fact Sheet Energies of prominent radiations (MeV) (Appendix A):

Alpha 5.49 Gamma 0.060

Likely exposure mode:

Inhalation

Relative ease of determining whether an intake of 1 CDG has occurred:

Difficult

A GM survey meter is likely not to be capable of showing the presence of external alpha contamination unless it has a thin detector window (pancake probe) and the survey is performed very slowly [~3 to 5 cm s–1 (1 to 2 inches s–1)] and within 0.5 to 1 cm (0.25 to 0.5 inches) of the surface. Inhalation of 241Am at the 1 CDG level will not likely be detectable using this method. Early nasal swabs should be collected for possible laboratory counting. A patient with an intake of 1 CDG is not likely to show external contamination following decontamination activities. Urine sampling is needed for subsequent radiochemistry laboratory analysis. Internally-deposited 241 Am requires highly specialized detection equipment for direct measurement. The typical method for detection is a urine sample for screening followed by large-volume urine and total fecal samples for dose assessment. Basis for Considering Therapy: Indication of internal contamination by wound or being close to the site of an RDD or IND incident for a possible inhalation exposure.

7.4 CESIUM-137: CLINICAL DECISION GUIDE FACT SHEET

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Therapy for ingestion is not indicated due to very low GI tract absorption of americium. Material form, exposure route:

Type M, inhalation

CDG:

0.0093 MBq (0.25 μCi)

Expected CDG Indicators: External contamination:

High levels of external contamination, may be readily reduced with decontamination

Nasal swabs shortly after inhalation:

2.8 × 104 dpm

Whole-body activity at 24 h:

Not applicable

Lung activity at 24 h:

3.2 × 104 dpm

Total urinary excretion (first 24 h):

1.0 × 103 dpm

Single-void urine during first 24 h:

See Section 7.2.3

7.4 Cesium-137: Clinical Decision Guide Fact Sheet Energies of prominent radiations with 137m Ba (MeV) (Appendix A):

Beta 1.18 Gamma 0.662

Likely exposure mode:

Inhalation or ingestion

Relative ease of determining whether an intake of 1 CDG has occurred:

Easy

After an intake of 137Cs at the 1 CDG level, the gamma emissions should be readily detectable with a GM survey meter. Inhalation at the 1 CDG level will likely present as detectable activity using a survey meter on the chest or abdomen. Early nasal swabs should show readily detectable activity. Survey-meter detection may predominate in the lung or GI tract regions. Basis for Considering Therapy: Indication of an internal deposition (e.g., continued detection of general internal deposition) at the whole-body levels below.

80 / 7. CLINICAL DECISION GUIDE: CONCEPT AND USE

Material form, exposure route:

Type F, inhalation

Soluble, ingestion

CDG:

58 MBq (1,600 μCi)

28 MBq (760 μCi)

Expected CDG Indicators: External contamination:

High levels of external contamination, may be readily reduced with decontamination

Nasal swabs shortly after inhalation:

1.7 × 108 dpm

Not applicable

Whole-body activity at 24 h:

2.0 × 109 dpm

1.6 × 109 dpm

Lung activity at 24 h:

Not applicable, clears rapidly

Not applicable

Total urinary excretion (first 24 h):

7.7 × 107 dpm

7.6 × 107 dpm

Single-void urine during first 24 h:

See Section 7.2.3

7.5 Cobalt-60: Clinical Decision Guide Fact Sheet Energies of prominent radiations (MeV) (Appendix A):

Beta 2.51 Gammas 1.17, 1.33

Likely exposure mode:

Inhalation, wound

Relative ease of determining whether an intake of 1 CDG has occurred:

Easy

After an intake of 60Co at the 1 CDG level, the high-energy gamma emission from activity inside the body should be readily detectable with a GM survey meter and will not decrease significantly following external decontamination. Inhalation or ingestion at the 1 CDG level will likely present as detectable activity using a survey meter everywhere on the torso. Early nasal swabs should show readily detectable activity. Even after extensive external decontamination efforts, 60Co in a patient with an intake of 1 CDG will be detectable several feet away from the meter.

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Basis for Considering Therapy: Indication of an internal deposition (e.g., continued detection of general internal deposition) at the whole-body and lung-activity indicator levels below. Material form, exposure route:

Type M, inhalation

Type S, inhalation

CDG:

35 MBq (950 μCi)

15 MBq (400 μCi)

Expected CDG Indicators:4 External contamination:

High levels of external contamination may be readily reduced with decontamination

Nasal swabs shortly after inhalation:

1.1 × 108 dpm

4.4 × 107 dpm

Whole-body activity at 24 h:

1.0 × 109 dpm

4.3 × 108 dpm

Lung activity at 24 h:

1.2 × 108 dpm

5.6 × 107 dpm

Total urinary excretion (first 24 h):

4.2 × 107 dpm

Uncertain4

Single-void urine during first 24 h:

See Section 7.2.3

4For this case, 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 given in Section 20 of NCRP Report 161 (NCRP, 2008a).

82 / 7. CLINICAL DECISION GUIDE: CONCEPT AND USE 7.6 Iodine-131: Clinical Decision Guide Fact Sheet Energies of prominent radiations (MeV) (Appendix A):

Beta 0.807 Gammas 0.0298, 0.364

Likely exposure mode:

Any (inhalation, ingestion, wound, absorption through intact skin)

Relative ease of determining whether an intake of 1 CDG has occurred:

Easy

Following a CDG intake, the medium energy beta and gamma emissions from activity inside the thyroid will likely be readily detectable using a GM survey meter placed over the thyroid. Ensure patient is externally decontaminated, with thyroid activity not significantly decreasing following external decontamination. Early nasal swabs should show readily detectable activity. Basis for Considering Therapy: Indication of an internal deposition (e.g., continued detection of general internal deposition) at the thyroid or whole-body activity indicator levels below. In cases where intake exceeding a CDG is suspected, it may be better to treat with KI rather than wait for confirmation by bioassay. Material form, exposure route:

Vapor, soluble intake

Clinical Decision Guide: The CDG values related to the FDA guidance and their associated bioassay excretion and retention values are given in Table 7.6. Thyroid blockade by administration of KI can substantially reduce uptake of radioactive iodine if administered 1 to 2 d before exposure or within 4 to 5 h after exposure. 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, to adults 18 to 40 y of age if the projected dose is ≥0.1 Gy, and to pregnant or lactating women or persons <18 y of age if the projected dose is ≥0.05 Gy. In this Report, CDGs for 131I (the only radioisotope 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

1.4 × 103 1.4 × 103 1.4 × 103 2.3 × 103 3.5 × 103 7.0 × 103 1.2 × 104 1.2 × 104

3.9 × 10–7 3.9 × 10–7 3.9 × 10–7 6.2 × 10–7 9.5 × 10–7 1.9 × 10–6 3.2 × 10–6 3.3 × 10–6

Adult >40 y

Adult 18 – 40 y

Pregnancy or lactation

Age 12 – 18 y

Age 7 – 12 y

Age 3 – 7 y

Age 0.5 – 3 y

Age <0.5 y

1.5 × 104

1.6 × 104

2.6 × 104

5.3 × 104

8.1 × 104

1.3 × 105

2.6 × 105

1.3 × 107

Bq

4.1 × 10–1

4.2 × 10–1

7.1 × 10–1

1.4 × 100

2.2 × 100

3.5 × 100

6.9 × 100

3.5 × 102

μCi

CDG (intake activity)

56

56

56

56

56

56

56

56

Urinary Excretion 0 – 24 h

22

22

23

23

23

23

23

23

Retention in Thyroid at 24 h

29

32

33

33

33

33

33

33

TotalBody Retention at 24 h

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

5.1 × 105

5.3 × 105

8.8 × 105

1.8 × 106

2.7 × 106

4.3 × 106

8.6 × 106

4.3 × 108

Urinary Excretion 0 – 24 h

2.0 × 105

2.1 × 105

3.6 × 105

7.3 × 105

1.1 × 106

1.8 × 106

3.5 × 106

1.8 × 108

Retention in Thyroid at 24 h

2.6 × 105

3.0 × 105

5.2 × 105

1.0 × 106

1.6 × 106

2.5 × 106

5.1 × 106

2.5 × 108

Total-Body Retention at 24 h

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

aThe tabulated values are based on threshold doses estimated by FDA (2001) for these different risk groups, together with age-specific biokinetic and dose estimates for 131I inhaled as a vapor listed in Table 20.56 in NCRP Report No. 161 (NCRP, 2008a). b Excretion and retention values are based on the assumption that the thyroid has not been blocked by the administration of stable iodine.

mrem μCi–1

Sv Bq–1

Group

Committed Equivalent Dose to Thyroid

TABLE 7.6—Model predictions used to assess whether an intake of 131I by inhalation as a vapor or ingestion in a soluble form exceeds the CDG.a 7.6 IODINE-131: CLINICAL DECISION GUIDE FACT SHEET

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84 / 7. CLINICAL DECISION GUIDE: CONCEPT AND USE unit intake of radioiodine given in NCRP Report No. 161 (NCRP, 2008a): adults >40 y of age; adults 18 to 40 y of age; pregnant or lactating women; and groups 12 to 18, 7 to 12, 3 to 7, 0.5 to 3, and <0.5 y of age. 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 groups 12 to 18, 7 to 12, 3 to 7, 0.5 to 3, and <0.5 y of age, 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 as given in Table 20.56 of NCRP Report No. 161 (NCRP, 2008a). Single-void urine during first 24 h:

See Section 7.2.3

7.7 Iridium-192: Clinical Decision Guide Fact Sheet Energies of prominent radiations (MeV) (Appendix A):

Beta 1.38 Gamma 0.316

Likely exposure mode:

Inhalation, wound

Relative ease of determining whether an intake of 1 CDG has occurred:

Medium

After an intake of 192Ir at the 1 CDG level, gamma should be readily detectable with a GM survey meter. Inhalation at the 1 CDG level will likely present as detectable activity using a survey meter on the chest or abdomen. Early nasal swabs should show readily detectable activity. Survey-meter detection may predominate in the lung or GI tract region. Basis for Considering Therapy: Indication of an internal deposition (e.g., continued detection of general internal deposition) at the whole-body indicator levels below. Material form, exposure route:

Type M, inhalation

Type S, inhalation

CDG:

59 MBq (1,600 µCi)

50 MBq (1,400 µCi)

7.8 PLUTONIUM-238: CLINICAL DECISION GUIDE FACT SHEET

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Expected CDG Indicators: External contamination:

High levels of external contamination; may be readily reduced with decontamination.

Nasal swabs shortly after inhalation:

1.8 × 108 dpm

1.5 × 108 dpm

Whole-body activity at 24 h:

1.7 × 109 dpm

1.5 × 109 dpm

Lung activity at 24 h:

2.0 × 108 dpm

1.9 × 108 dpm

Total urinary excretion (first 24 h) after an inhalation exposure:

1.1 × 107 dpm

1.1 × 106 dpm

Single-void urine during first 24 h:

See Section 7.2.3

7.8 Plutonium-238: Clinical Decision Guide Fact Sheet Energies of prominent radiations (MeV) (Appendix A):

Alpha 5.46; 5.50 X rays 0.0173

Alpha activity poses difficulties for detection by GM survey meter. Very low-energy x rays not capable of detection by GM survey meter. Likely exposure mode:

Inhalation or wound

Relative ease of determining whether an intake of 1 CDG has occurred:

Very difficult

A GM survey meter is likely not to be capable of showing the presence of alpha contamination unless it has a thin detector window (pancake probe) and the survey is performed very slowly [~3 to 5 cm s–1 (1 to 2 inches s–1)] and within 0.5 to 1 cm (0.25 to 0.5 inches) of the surface. Inhalation at the CDG level will not likely be detectable using this method. Early nasal swabs should be collected for laboratory counting. A patient with an intake of 1 CDG is not likely to show external contamination following decontamination activities. Urine sampling is needed for subsequent analysis in a radiochemistry laboratory. internally-deposited plutonium requires highly specialized detection equipment for direct measurement. The typical method for detection is a urine sample for screening, followed by large-volume urine and total fecal samples for dose assessment.

86 / 7. CLINICAL DECISION GUIDE: CONCEPT AND USE Basis for Considering Therapy: Indication of an internal deposition by wound or close proximity to an incident for a possible inhalation exposure. Therapy for ingestion is not indicated due to very low GI tract absorption of plutonium. Material form, exposure route:

Type M, inhalation

Type S, inhalation

CDG:

0.0081 MBq (0.22 μCi)

0.023 MBq (0.61 μCi)

Expected CDG Indicators: External contamination:

May not be detectable using GM survey meter

Nasal swabs shortly after inhalation:

2.4 × 104 dpm

Whole-body activity at 24 h:

Not relevant for measurement

Lung activity at 24 h:

2.8 × 104 dpm

8.7 × 104 dpm

Total urinary excretion (first 24 h):

1.0 × 102 dpm

2.9 × 100 dpm

Single-void urine during first 24 h:

See Section 7.2.3

6.8 × 104 dpm

7.9 Plutonium-239: Clinical Decision Guide Fact Sheet Energies of prominent radiations (MeV) (Appendix A):

Alpha 5.18, X ray 0.0136 and others

Alpha activity poses difficulties for detection by GM survey meter. Very low-energy x rays not capable of detection by a GM survey meter. Likely exposure mode:

Inhalation or wound

Relative ease of determining whether an intake of 1 CDG has occurred:

Very difficult

7.9 PLUTONIUM-239: CLINICAL DECISION GUIDE FACT SHEET

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A GM survey meter is likely not to be capable of showing the presence of alpha contamination unless it has a thin detector window (pancake probe) and the survey is performed very slowly [~3 to 5 cm s–1 (1 to 2 inches s–1)] and within 0.5 to 1 cm (0.25 to 0.5 inches) of the surface. Inhalation at the CDG level will not likely be detectable using this method. Early nasal swabs should be collected for laboratory counting. A patient with an intake of 1 CDG is not likely to show external contamination following decontamination activities. Urine sampling is needed for subsequent analysis in a radiochemistry laboratory. Internally-deposited plutonium requires highly specialized detection equipment for direct measurement. The typical method for detection is a urine sample for screening, followed by large-volume urine and total fecal samples for dose assessment. Basis for Considering Therapy: Indication of an internal deposition by wound or close proximity to an incident for a possible inhalation exposure. Therapy for ingestion is not indicated due to very low GI tract absorption of plutonium. Material form, exposure route:

Type M, Inhalation

Type S, Inhalation

CDG:

0.0076 MBq (0.20 μCi)

0.030 MBq (0.80 μCi)

Expected CDG Indicators: External contamination:

May not be detectable using GM survey meter

Nasal swabs shortly after inhalation:

2.3 × 104 dpm

Whole-body activity at 24 h:

Not relevant for measurement

Lung activity at 24 h:

2.6 × 104 dpm

1.1 × 105 dpm

Total urinary excretion (first 24 h):

9.6 × 101 dpm

3.8 × 100 dpm

Single-void urine during first 24 h:

See Section 7.2.3

8.9 × 104 dpm

88 / 7. CLINICAL DECISION GUIDE: CONCEPT AND USE 7.10 Radium-226: Clinical Decision Guide Fact Sheet Energies of prominent radiations (MeV) (Appendix A):

Alphas 4.78 and others Gammas 0.186 and others

Likely exposure mode:

Inhalation

Relative ease of determining whether an intake of 1 CDG has occurred:

Difficult

After an intake of 226Ra in equilibrium with its progeny at the 1 CDG level, the alpha and beta emissions from activity inside the body will likely be completely shielded by the body tissue and distance. It is possible that some gamma emissions might be detected but recognizing a difference from background using a GM survey meter with pancake probe will probably not be possible. Early nasal swabs showing readily detectable activity would support an inhalation, and should be retained for possible laboratory counting. Urine sampling is needed for subsequent radiochemistry laboratory analysis. Internally-deposited 226Ra requires specialized detection equipment (whole-body counter) for direct measurement. The typical method for detection is a urine sample for screening followed by a large-volume urine sample. Basis for Considering Therapy: Indication of internal contamination by wound or being close to the site of an RDD or IND incident for a possible inhalation exposure. Material form, exposure route:

Type M, inhalation

CDG:

0.11 MBq (3.1 μCi)

Expected CDG Indicators: External contamination:

High levels of external contamination, may be readily reduced with decontamination

Nasal swabs shortly after inhalation:

3.4 × 105 dpm

Whole-body activity at 24 h:

3.4 × 106 dpm

Lung activity at 24 h:

4.0 × 104 dpm

Total urinary excretion (first 24 h):

1.1 × 104 dpm

Single-void urine during first 24 h:

See Section 7.2.3

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7.11 Strontium-90: Clinical Decision Guide Fact Sheet Energies of prominent radiations with 90 Y (MeV) (Appendix A):

Betas 0.546, 2.28 No Gammas

Likely exposure mode:

Inhalation or ingestion

Relative ease of determining whether an intake of 1 CDG has occurred:

Difficult

After an intake of 90Sr at the 1 CDG level, the beta emissions from activity inside the body will likely be significantly shielded by the body tissue and distance. Some indication of internal contamination might be noted on a GM survey meter if the observed counts do not decrease significantly after external decontamination of the body. Inhalation or ingestion at the 1 CDG level will likely present as some detectable activity using a survey meter on the abdomen. Early nasal swabs should show readily detectable activity after an inhalation exposure. Survey meter detection may predominate in the GI tract region following removal of external contamination. Basis for Considering Therapy: Indication of an internal deposition (e.g., continued detection of general internal deposition) at whole-body indicator levels below. Whole-body counting will require equipment capable of detecting bremsstrahlung radiations. Material form, exposure route:

Type F, inhalation

Soluble, ingestion

CDG:

8.3 MBq (230 μCi)

8.9 MBq (240 μCi)

Expected CDG Indicators: External contamination:

High levels of external contamination, may be readily reduced with decontamination

Nasal swabs shortly after inhalation:

2.5 × 107 dpm

Not applicable

Whole-body activity at 24 h:

2.5 × 108 dpm

3.9 × 108 dpm

Lung activity at 24 h:

Not applicable

Not applicable

90 / 7. CLINICAL DECISION GUIDE: CONCEPT AND USE

Total urinary excretion (first 24 h):

3.4 × 107 dpm

3.0 × 107 dpm

Single-void urine during first 24 h:

See Section 7.2.3

7.12 Uranium: Clinical Decision Guide and Nephrotoxicity Fact Sheet 7.12.1 Treatment Based on Radiological Properties of Uranium (CDG) As described in Section 7.2.6, internally-deposited uranium can pose both radiological and chemical (nephrotoxic) risks depending on the intake level, in vivo solubility, and 235U enrichment. The CDG facts in this section relate to exposures for which radiological toxicity is the limiting case as illustrated in Figure 7.2. These values, like the CDG values for other radionuclides in this section were taken directly from Tables 11.1 and 11.2 of NCRP Report No. 161 (NCRP, 2008a). Additional guidance based on chemical toxicity is given below in Section 7.12.2. Commonly found as mixtures of 238U + 234U + 235U in natural uranium or DU: Radiation type:

Alpha, beta and gamma radiations from uranium, but also likely to have beta and gamma radiations from progeny unless freshly processed

Likely exposure mode:

Inhalation (Type M or S)

Relative ease of determining whether an intake of 1 CDG has occurred:

Medium to difficult

See Section 7.2.6 for factors determining the limiting case for radiological and chemical toxicity for inhalation of different types of uranium having different solubility types. Basis for Considering Therapy: Figure 7.2 and Table 7.5 provide needed information related to the fraction of 235U in the uranium, the absorption type, and the limiting CDG (i.e., radiological or chemical toxicity).

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Radiological CDG Information for Natural Uranium (specific activity = 0.7 μCi g–1 uranium): Material form, exposure route:

Type M, inhalation

Type S, inhalation

CDG:

0.12 MBq (3.2 μCi) (4.6 g)

0.037 MBq (1.0 μCi) (1.4 g)

Expected CDG Indicators: External contamination:

High levels of external contamination, may be readily reduced with decontamination

Nasal swabs shortly after inhalation:

3.6 × 105 dpm

1.1 × 105 dpm

Whole-body activity at 24 h:

Not applicable

Not applicable

Lung activity at 24 h:

4.2 × 105 dpm

1.4 × 105 dpm

Total urinary excretion (first 24 h):

1.6 × 105 dpm

1.5 × 103 dpm

Single-void urine during first 24 h:

See Section 7.2.3

7.12.2 Treatment Based on Nephrotoxic Properties of Uranium As can be seen in Figure 7.2, there are certain exposure scenarios involving uranium of different in vivo solubility and 235U enrichment for which the chemical nephrotoxic threshold will be the limiting case, especially for Type F forms of uranium. The CDG information provided in NCRP Report No. 161 (NCRP, 2008a) noted that there are cases where a chemical threshold would be the limiting case but CDG values based on chemical nephrotoxicity were not provided. Development of specific CDGs for uranium based on its nephrotoxicity is beyond the scope of this Report. However, Table 7.5 provides a valuable resource for a physician needing to determine the level at which treatment for the nephrotoxic properties of uranium should be considered. As a point of reference, occupational exposure levels are based on a nephrotoxic level of 3 µg g–1 of uranium in kidney. The three levels of uranium intake in the body corresponding to the horizontal lines based on this concentration limit for the three lung absorption types are 0.029 g for

92 / 7. CLINICAL DECISION GUIDE: CONCEPT AND USE Type Chem F, 0.23 for Type Chem M, and 7.4 g for Type Chem S. In the absence of further guidance from other sources, the physician can select a kidney uranium concentration from Table 7.5 as being a value at which treatment for a uranium intake should be considered and this value can be used to determine the corresponding measured indicators. For example, if the physician chose to use the 3 µg g–1 of uranium in kidney uranium concentration as a trigger level for considering treatment, Table 7.5 shows that this level falls within the renal effects Group 1 category that may result in transient indications of renal dysfunction, but the subject is not likely to become ill. The following information provides measured values based on a peak uranium concentration in the kidney of 3 µg g–1 of uranium in kidney. If a different peak concentration is chosen as the level at which therapy should be considered, the magnitudes of the listed measured values should be raised or lowered in direct proportion to the level chosen. For instance, if a peak uranium concentration of 5 µg g–1 of uranium in kidney was chosen as a possible trigger for considering therapy, the measured values listed below should be increased by 5 ÷ 3 = 1.67. Nephrotoxicity-related information for natural uranium (specific activity = 0.7 μCi g–1 uranium) based on a peak kidney concentration of 3 µg g–1 of uranium in kidney: Material form, exposure route:

Type F, inhalation

Type M, inhalation

Uranium intake:

7.5 × 10–4 MBq (0.02 μCi) (0.029 g)

6.0 × 10–3 MBq (0.16 Ci) (0.23 g)

Nasal swabs shortly after inhalation:

Not applicable

1.8 × 104 dpm

Whole-body activity at 24 h:

Not applicable

Not applicable

Lung activity at 24 h:

Not applicable

2.1 × 104 dpm

Total urinary excretion (first 24 h):

8.2 × 103 dpm

8.3 × 103 dpm

Single-void urine during first 24 h:

See Section 7.2.3

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7.13 Clinical Decision Guide Technical Details 7.13.1 Noniodine Radionuclides For radionuclides other than isotopes of iodine, the CDG is the maximum, once-in-a lifetime intake of a radionuclide that: (1) represents a stochastic risk, as judged by the calculated effective dose over 50 y of age for intake by adults and to 70 y of age for intake by children, that is in the range of risks associated with guidance on dose limits for emergency situations (DOE, 2009b; FEMA, 2008b; ICRP, 1991; 2005; NCRP, 1993; 2005); and (2) does not cause deterministic effects as judged by the calculated 30 d RBE-weighted absorbed doses to red marrow and lungs, with allowance for uncertainties typically involved in the dose estimates. For all radionuclides other than the isotopes of iodine, the CDG doses in adults are 0.25 Sv (25 rem) (50 y committed effective dose) for consideration of late-occurring cancers; a 30 d RBE-weighted absorbed dose value of 0.25 Gy-Eq (25 rad-Eq) for consideration of bone marrow damage; and a 30 d RBE-weighted absorbed dose value of 1 Gy-Eq (100 rad-Eq) for consideration of pulmonary damage. For intake of a radionuclide other than an isotope of iodine, the CDG for an adult is the maximum intake satisfying all three of these dose constraints based on tissue damage and cancer as shown in Equation 7.1. 0.25 Gy – Eq 0.23 Sv CDG = MIN ------------------------------- , -------------------------------------------------------------------- , –1 –1 e ( Sv Bq ) d Red Marrow ( Gy – EqBq )

(7.1)

1.0 ----------------------------------------------------–1 d Lung ( Gy – EqBq ) 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 meets the constraint on the effective dose and the 30 d absorbed doses to the red marrow and lungs. With the following exceptions relating to 192Ir, the CDGs for radionuclides addressed in this Report are determined by the estimated risk of stochastic effects. The CDG for 192Ir inhaled in an either moderately soluble (Type M) form or a relatively-insoluble (Type S) form is determined by the 30 d RBE-weighted absorbed

94 / 7. CLINICAL DECISION GUIDE: CONCEPT AND USE dose to the lungs (NCRP, 2008a). Most CDGs tabulated in this Report are for inhaled radionuclides, but intake by ingestion is also considered for some radionuclides when appropriate. As discussed in Section 7.1, the CDGs for radionuclides other than isotopes of iodine for children (0 to 18 y of age) and pregnant women 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. Table 7.1 gives CDG values for the different exposure routes and solubility types considered in this Report. More CDG information on these radionuclides is given in the CDG fact sheets in Sections 7.3 to 7.12. 7.13.2 Iodine Radionuclides Thyroid blockade by administration of KI can substantially reduce uptake of radioactive iodine by the thyroid if administered 1 to 2 d before exposure or within 4 to 5 h after exposure. 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 the thyroid is ≥5 Gy, to adults 18 to 40 y of age if the projected dose is ≥0.1 Gy, and to pregnant or lactating women or persons <18 y of age if the projected dose is ≥0.05 Gy. In this Report, CDGs for 131I (the only radionuclide 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 given in Section 20 of NCRP Report No. 161 (NCRP, 2008a): adults >40 y of age; adults 18 to 40 y of age; pregnant or lactating women; and groups 12 to 18, 7 to 12, 3 to 7, 0.5 to 3, and <0.5 y of age. 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 groups 12 to 18, 7 to 12, 3 to 7, 0.5 to 3, and <0.5 y of age, 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 as given in Table 20.56 of NCRP Report No. 161 (NCRP, 2008a). Further information is provided in Section 7.6. 7.13.3 Uranium Mass and Activity Calculations As shown in Section 7.2.6 and 7.12, treatment decisions for removal of internally-deposited uranium can be triggered by either the activity or mass associated with this deposition, depending of

7.13 CLINICAL DECISION GUIDE TECHNICAL DETAILS

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the characteristics of the exposure material. Sometimes it will be necessary to convert a sample result (like a urine sample measurement) from mass in gram of uranium to activity in becquerel or μCi or vice versa for comparison with the listed indicator values given in Sections 7.12.1 and 7.12.2. The specific activity of the particular uranium material, given in Bq g–1 uranium or μCi g–1 uranium is used for this conversion. In the report by Hinnefeld et al. (1988), the specific activity for DU is ~0.4 μCi g–1 uranium and for natural uranium, the specific activity is ~0.7 μCi g–1 uranium. For uranium with other characteristics, use one of the following equations based on the uranium enrichment (UE) expressed as percent 235U by mass based on information in (Hinnefeld et al., 1988): specific activity = 4

(7.2) 2

= ( 3.7 × 10 ) ( 0.4 + 0.38 UE + 0.0034 UE ) Bq g

2

–1

specific activity = ( 0.4 + 0.38 UE + 0.0034 UE ) μCi g

U,

–1

U.

(7.3)

8. Rapid Determination of Internal Contamination In the aftermath of an RDD or IND incident, casualties may be concerned about the radiation exposure they receive from external and internal radionuclide contamination, and healthcare providers may be concerned about exposures they could receive while treating patients. These issues are especially important in the event of mass casualties from an IND or large RDD incident because of the need to rapidly assess and treat severe injuries, especially when it is necessary to delay the decontamination process to address the immediate health and safety needs of some patients (Smith et al., 2005). Therefore, it is important to conduct radiological triage of patients who are potentially contaminated internally while at the same time addressing medical triage of injured patients. Section 6 addresses radiological triage and screening of patients at the scene and at the hospital and provides flow diagrams for triage and screening of patients. Once external contamination issues have been addressed, casualties should be screened for internal contamination either at the scene or at the hospital. But in no case should medical treatment be delayed for severe injuries or medical conditions, even when patients have not been surveyed for external contamination or screened for internal contamination. This section addresses screening for internal contamination using a variety of radiation detection instrumentation, but focuses on use of the more commonly available GM survey meter for rapid screening of patients. In an incident involving a small number of patients, each patient should be decontaminated and then screened with an appropriate instrument available at the scene or hospital. DHHS (2009) provides an algorithm for triage of contaminated patients. If the incident causes mass casualties, emergency responders at the scene and triage personnel at the hospital should step up their responses (Section 12) and may need to conduct a quick screen to determine by interview whether patients were close enough to the point of release of radioactive materials to warrant screening with a survey instrument. In the event of an RDD incident, it should be assumed that patients who were close enough to the incident to have sustained injuries may be contaminated internally and should be screened with a survey instrument. Patients without external 96

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contamination, provided they have not been decontaminated elsewhere, will not be contaminated internally and will not need to be screened for internal contamination. Patients may be interviewed to determine their location relative to the explosion and will not need to be screened if located up wind at the time of the incident. The goal of a screening program is to survey all individuals who could be internally contaminated. However, in an incident involving the possible contamination of hundreds or thousands, screening resources may not be adequate to permit prompt screening of all people. A possible strategy to determine which group(s) of people may be most likely to have internal contamination would be to screen a few people from each of various groups (e.g., highly contaminated people) who were outdoors and close to the device at the time of the explosion versus people who passed through the plume downstream of the incident (Harper et al., 2007) versus people who were inside buildings at the time of the release. The information gathered on the likely amounts of internal contamination of people in each group would be useful in guiding the decision on whom to screen first. 8.1 Rapid Identification of Radionuclide(s) Involved Rapid identification of the radionuclide(s) involved in a radiological contamination incident is essential to the selection of appropriate methods for assessment of internal contamination and subsequent treatment decisions. Rapid identification of the radionuclide(s) involved in an incident may be performed by several means. For example, the identity of radionuclide(s) in widespread contamination resulting from a transportation accident could be accomplished by a review of the shipping papers. In other instances, it may not be immediately obvious that radioactive material is involved in an incident and its presence may not be detected until a radiation and contamination survey is performed. In some localities, real-time environmental monitoring may reveal the presence of radioactive material, and radionuclide identification may be provided by a state laboratory. Many hazardous materials response teams and some facilities are now equipped with portable multichannel analyzer instruments for radionuclide identification. These portable instruments typically incorporate a small, solid-state detector and electronics that allow the generation of a gamma-ray spectrum. The devices are programmed to identify commonly-encountered radionuclides such as 99m Tc, 131I, 60Co, and others. Most are also equipped to display dose rate, usually in units of μGy h–1 or μR h–1.

98 / 8. RAPID DETERMINATION OF INTERNAL CONTAMINATION

The rapid detection and identification of pure alpha- or beta-emitting radionuclides (e.g., 210Po, 90Sr/90Y) are challenging and should be addressed in the planning process. Most beta-emitting radionuclides can be detected with commonly used radiation survey instruments, but the identification of a specific radionuclide may require radiochemical laboratory analysis. Some high-energy beta emitters (e.g., 90Sr/90Y) may appear as a continuous spectrum without discrete photopeaks (bremsstrahlung continuum) on portable multichannel analyzers if the levels are sufficiently large. The presence of such a broad continuum without the sharply defined photopeaks from typical gamma-emitting radionuclides could be interpreted as a pure beta-emitting radionuclide. 8.2 Screening for External Contamination Screening for external contamination on persons involved in a radionuclide contamination incident is necessary to determine the need for decontamination and to identify radioactive material that could be misidentified as internal contamination in subsequent assessments. In some cases, external decontamination may be incomplete or impractical because of the need to avoid damaging the skin. In such cases, it is important to have an accurate characterization of residual external contamination so that this may be taken into account when performing assessments of internallydeposited radionuclides. In all but the most extreme cases, standard precautions provide adequate protection to healthcare workers to prevent secondary contamination. The presence of external contamination should almost never delay urgent medical care. External contamination rarely, if ever, represents an acute radiological hazard to the patient or healthcare providers (Goans, 2004; Mettler and Voelz, 2002; NCRP, 2005; Smith et al., 2005). Studies have shown that when the outer clothing of the contaminated individual is removed, ~90 % of the radioactive contamination is eliminated (Miller et al., 2005; NCRP, 2005). It is prudent, however, to use a survey meter to assess any patient who has not received decontamination procedures due to severe injury. Decontamination should not interfere with, or take precedence over, medical care of patients with life-threatening injuries or

8.3 DIRECT SCREENING FOR INTERNAL CONTAMINATION

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illness (Bartlett and Greenberg, 2005; HSC, 2009). An RDD or IND could possibly embed discrete fragments of radioactive material (e.g., shrapnel) into the patient (HSC, 2009; Smith et al., 2005). Shielding should be considered if dose rates emanating from the patient are in excess of 1 mGy h–1 (100 mrad h–1) at 30 cm (1 foot) to limit exposure to healthcare personnel. Prompt debridement or excision to remove such discrete fragments will limit both patient and caregiver doses. A shielded container, such as those commonly found in a nuclear medicine department, should be used to contain and store such radioactive debris. Table 8.1 presents a summary of instruments appropriate for personnel contamination surveys. 8.3 Direct (in vivo) Screening for Internal Contamination 8.3.1

Detection of Internal Contamination by Direct Measurement

The direct detection and measurement of internally-deposited radionuclides is a complex process (IAEA, 1996; ICRU, 2003), that will be significantly complicated in a large-scale contamination incident by the possibility of external contamination being interpreted as internal contamination (or vice versa). It is important to recognize the difference between the detection of radioactive material and the accurate quantification of intakes. The methods recommended in this section are for the rapid screening of large numbers of potentially-contaminated patients only and should not be used as a basis for definitive dose assessment. These methods and procedures are intended only to rapidly identify patients who (1) may benefit from medical intervention to reduce dose, and (2) to identify patients who should receive follow-up measurements and assessment (Figure 8.1). For purposes of this section, it is presumed that external radioactive contamination was removed during decontamination by emergency responders (e.g., fire department) or by hospital personnel and that any radiation detected is the result of internal contamination. However, the possibility of persistent external contamination should be considered. In the case of gamma-emitting radionuclides, it may be possible to distinguish persistent external contamination from internally-deposited contamination by surveying the entire body of the patient. Any difference between the anterior and posterior survey results is a crude indicator of persistent external contamination. Also, high survey readings on face, hands and feet compared to covered surfaces may indicate persistent external contamination and the need for further decontamination prior to screening for internal contamination.

Ionization chamber, micro-R meter NaI(Tl) scintillation (internally mounted)

Dose rate

Portable multichannel analyzer

Dose rate

mGy (mR) h–1 or μGy (μR) h–1

Radionuclide identification based on gamma energy

Dose rate (gamma or x rays only) contamination

mGy (mR) h–1 cpm

Various – radionuclide identification, counts per minute

Dose rate contamination

mGy (mR) h–1 cpm

Also see Becker et al. (1991) regarding instrument usage in an emergency situation.

NaI(Tl) scintillation

General survey instrument

a

Energy-compensated GM

General survey instrument (rate meter)

Beta-gamma contamination on hands and feet when leaving a room or area

Alpha contamination

Beta-gamma contamination

Use for These Surveys

counts per minute

counts per minute

Alpha scintillation Thin-window GM pancake

counts per minute

Instrument Read-Out Units

Thin-window GM pancake

Probe

Hand and foot monitor

General survey instrument (rate meter)

Instrument Type

TABLE 8.1—Radiation instrumentation for personnel contamination surveys.a

100 / 8. RAPID DETERMINATION OF INTERNAL CONTAMINATION

8.3 DIRECT SCREENING FOR INTERNAL CONTAMINATION

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Fig. 8.1. Screening process to determine whether a patient may contain internal contamination (adapted from NCRP, 2008a).

8.3.1.1 GM Survey Meter. GM survey meters are used commonly to detect external radioactive contamination on personnel and work surfaces in hospitals, university research laboratories, and other work places where radioactive materials are used. Because they are commonly available in hospitals, a discussion of their use is provided below in Section 8.3.2.

8.3.1.2 Whole-Body and Lung Counters. For occupational monitoring purposes, whole-body counters are used to detect and quantitate internally-deposited radionuclides that emit photons with

102 / 8. RAPID DETERMINATION OF INTERNAL CONTAMINATION energies in excess of ~200 keV. Whole-body counters use gamma scintillation (e.g., sodium iodide) or semiconductor (e.g., germanium) photon detectors in a low-background environment achieved through specially constructed, shielded counting cells or shadow shields. Counting configurations require subjects to stand, sit in a constant geometry chair, or recline on a fixed or moving bed. A standing configuration can be used to count the subject in a relatively short time (e.g., 3 min) but is correspondingly less sensitive than a sitting or reclining configuration for which the counting interval may run 10 to 20 min. However, this short counting time makes the standing whole-body counter a potential candidate for screening a population for internal contamination. Scanning whole-body counts (using a moving bed or moving detector configuration) may be capable of providing information about radionuclide distribution in the body, as well as quantifying the amount present. Whole-body counters are expensive to install and maintain and they are not widely available. However, they are present in large nuclear facilities such as power reactors, nuclear material processing facilities, national laboratories, and specialized research facilities. Some portable whole-body counters exist but they are limited in numbers and not likely to be available for immediate response in a scenario requiring screening of the public. They would need to be transported to the site of the incident, which would make them impractical for early screening purposes. Improvised whole-body counters, using a NaI(Tl) detection system with a fixed-chair geometry and portable lead shielding, as was done following the Goiânia accident (IAEA, 1998), could be used to quantitate internal contamination in subjects who had been identified by screening as containing internal contamination. Radionuclides that emit only low-energy photons (e.g., the 60 keV photon associated with 241Am as a separate radionuclide or as an ingrown progeny of 241Pu) are significantly more difficult to detect and quantitate. In an occupational environment, lung counting is used by some facilities to assess internal deposition of lowenergy photon-emitting radionuclides. Lung-counting systems typically use germanium, sodium iodide, or phoswich detectors in a heavily shielded room. The subject is placed in a sitting, reclining or supine position, and counting times may range from 20 min to an hour or more. Interpretation of results may require correction to give a more accurate estimate of the existing lung burden because the chest count can also include contributions from the skeleton (ribs, sternum and vertebrae). Without correction, the measured chest count may erroneously be assumed to represent a lung burden. Chest count results also can be highly dependent on chest-wall

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thickness, and chest-wall thickness is usually estimated for subjects by either a height-to-weight ratio or by measurement using ultrasound techniques. Portable lung-counter systems exist but are not likely to be readily available for emergency response. Considering these complications, use of lung counters to screen a population for internal contamination may not be practical. Direct bioassay using whole-body counters and lung counters could be viable for long-term follow-up of a limited number of exposure cases. Special arrangements for counting the subjects would be required, including the transport of the subject to the facility with the counting system or bringing a system to the subjects. However, such arrangements are likely to require notice and planning several days in advance of the measurements. Thus, these types of measurements are not considered viable for initial screening of populations, but may be suitable for secondary or tertiary measurements. A more detailed discussion of the capabilities of whole-body and lung counters is provided in NCRP Report No. 161 (NCRP, 2008a) and IAEA Safety Series Report No. 114 (IAEA, 1996). 8.3.2

Hospital Equipment for the Detection and Quantitation of Radionuclides

Several types of instruments available in some hospitals are capable of detecting a number of the radionuclides of concern in this Report (e.g., 60Co, 137Cs, and 192Ir) when deposited internally at a level corresponding to 1 CDG. The radionuclides 90Sr, 210Po, 226Ra, 234,235,238 U, 238,239Pu, and 241Am either have CDG values that are too small for this equipment to detect them or have radioactive emissions that cannot penetrate from inside the body to interact with the instrument. In such cases, alternative detection methods must be employed. Interviewing patients to ascertain their location during the incident and subsequent surveys for external contamination in the head and facial region can help determine potential intakes. If it is known that the easily detectable radionuclides (e.g., 60Co, 137 Cs, and 192Ir) are involved, the use of GM survey meters may be the best choice. These are readily available, easy to use, relatively inexpensive, and portable. Other equipment can be utilized but can be difficult to setup properly and use and may not be portable. In addition to the portable instrumentation described in Table 8.1, many hospitals have equipment that could be applied to population screening. Specific examples are described in the following sections. 8.3.2.1 GM Survey Meter. Most nuclear medicine departments use portable, hand-held GM survey meters to detect contamination on

104 / 8. RAPID DETERMINATION OF INTERNAL CONTAMINATION work surfaces or personnel. This instrument has the advantage of being relatively inexpensive and sensitive to most gamma, beta, and some alpha radiation, although its gamma detection efficiency is low. The detector contains a thin window with a surface area that varies from a pancake model (typically 15 cm2) to a standard endwindow probe (6 cm2), which affects the efficiency. A thin mica window retains the pressurized counting gas, usually a halogenquenched inert gas, inside the detector. Placing a thick cap over the window to minimize beta sensitivity is recommended when surveying for internal contamination from gamma-emitting radionuclides. Use of the GM detector will enable personnel to detect internal levels of 60Co, 137Cs, and 192Ir. It is not possible to measure 241Am internal contamination with a GM detector (Anigstein et al., 2005). GM detectors are energy dependent with the counting efficiency dependent on the gamma-ray energies and abundances. The count rate of GM detectors will vary with distance from radiation source, so when using this type of instrument, care should be used to keep the distance consistent. 8.3.2.2 Nuclear Medicine Thyroid-Uptake Probe. This probe is used in nuclear medicine mainly for measuring thyroid uptake of radioactive iodine and would be the preferred method for measuring 131I in the thyroid of subjects who had been exposed to 131I during an incident. Typical systems use a 50 × 50 mm (2 × 2 inches) cylindrical sodium iodide [NaI(Tl)] crystal coupled to a photomultiplier tube that is connected to a multichannel analyzer. Such a multichannel analyzer is now frequently constructed on a computer board with an associated library of radionuclides. Although the counting efficiency of this instrument is very good for its original intended use of detecting and quantifying radioiodine in the thyroid, this instrument has a rather low efficiency when used to detect whole-body internal contamination because of the shielding used around the detector. However, it may be useful as a lung counter by counting one lung at a time. Due to the attenuation of lowenergy gamma photons by the chest wall, the thyroid probe probably would not be capable of detecting a CDG of 241Am in the lungs. Test results (Anigstein et al., 2005) indicate that the system is useful for detecting surface contamination at activity levels of 37 kBq (1 μCi) or less of 60Co, 137Cs, 192Ir, and 241Am and at distances up to at least 60 cm (24 inches). The uptake probe is useful for detecting activity at depths at least midway into the body. The equipment often has an internal library that can be modified to include other radionuclides of interest.

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8.3.2.3 Portal Monitors. Since NRC issued Information Notice 91-03 (NRC, 1991), many hospitals have installed portal monitors to check trash and laundry for the presence of radioactive material before it leaves the hospital. These systems are somewhat portable and could be moved to another site in the hospital to scan people actively as they walk through the portal. The typical hospital system employs a pair of lead-shielded 75 × 75 mm (3 × 3 inches) NaI(Tl) detectors that face each other across a doorway. The detectors share a common digital alarming rate meter with printer output. The monitor records the time, date, and radiation level of a radioactive source when it crosses the portal threshold. In addition to the printer log, each alarm trip produces an audible alarm. If the patient has significant internal contamination, or there are too many contaminated people in a line of sight to the detectors, the background of the monitor may rise above the alarm threshold. Moving people away or trying to shield the monitor behind a concrete, masonry or lead-shielded wall may be necessary to reduce the background radiation to a level that reduces false positive alarms. These systems are typically capable of detecting external contamination at distances of 40 to 60 cm (16 to 24 inches). Furthermore, if the trigger level in µGy h–1 (μR h–1) is set low enough, it can detect 60Co, 137Cs, and 192Ir in activity levels of 370 kBq (10 μCi) or more at depths up to 25 cm in the body. These systems can detect 370 kBq (10 μCi) 241Am to about half this depth or up to ~12 cm deep (with the patient as close as possible to one of the detectors). The portal monitoring system is also able to detect 131I as either a point source or a diffuse source, and it appears that the system would be able to detect activity levels of the radionuclides tested at or below 37 kBq (1 μCi). Under normal circumstances, this system is operated as a two-detector system. The background as measured by each detector are ~0.015 μGy h–1 (1.5 μR h–1) or 0.03 μGy h–1 (3 μR h–1). Thus, an alarm level is usually set at three times background or 0.1 μGy h–1 (10 μR h–1) (Anigstein et al., 2005). The use of both of the portal monitor’s detectors provides greater sensitivity and allows patients to be directed through the doorway containing the detectors. The detectors will locate and detect the contamination in seconds. Even with having patients turning at 90 degree angles, only a few minutes per patient are required for detection. Most portal monitors are not equipped to identify specific radionuclides, so identity of the radionuclide should be determined by other means. One major issue that has been studied is the sensitivity of this instrument when measuring a heavily contaminated individual or a large population containing large quantities of radioactive

106 / 8. RAPID DETERMINATION OF INTERNAL CONTAMINATION contamination when considered together. In this case, the placement of the portal monitors becomes crucial. Experimentation has shown that in this scenario, an open-air area as large as a soccer stadium may not provide sufficient distance between detector and groups of people to prevent the alarming of the detectors (Kramer et al., 2006). Advanced planning for this contingency is necessary. Portal monitors must be placed behind sufficient shielding to prevent radiation from other sources, including patients, from interfering with measurements. 8.3.2.4 Nuclear Medicine Gamma Camera. Hospitals with nuclear medicine departments have nuclear medicine gamma cameras. Studies suggest that the efficiency of detection is reasonable and that this equipment can be used to detect internal contamination (Anigstein et al., 2007; Wallstrom et al., 1999). Gamma cameras detect radiation emanating from a patient’s body and use this information to create a projection image depicting the distribution of radioactive material. The images provide counting information pertaining to the radionuclide concentrations in individual organs. A gamma camera typically consists of a planar NaI(Tl) scintillation crystal coupled to a two-dimensional array of photomultiplier tubes. The crystal and photomultiplier tubes are housed in a camera-head housing. The scintillation crystal is typically 1 cm thick, which is optimal for imaging the 140 keV gamma ray from 99mTclabeled radiopharmaceuticals. However, some cameras have crystals as thick as 2.5 cm for imaging radionuclides that emit higher-energy photons. Cameras designed for imaging larger areas have larger dimensions. Higher-energy photons, such as those characteristic of 137Cs and 60Co, are more efficiently captured and detected by thicker crystals. Since most nuclear medicine diagnostic procedures use radionuclides that have lower-energy photons than those from 137Cs or 60C, the thinner crystal is the one more commonly used. Key components of gamma-camera systems are: • sodium-iodide [NaI(Tl)] detector that converts absorbed photon energy to scintillation pulses; • photomultiplier tubes that detect and amplify the scintillation pulses produced within the NaI(Tl) crystal; • pulse-height analyzers that allow for the discrimination specific gamma-ray energies; and • computers that determine exact locations within the NaI(Tl) crystal where scintillation incidents occurred and use this information to construct images of radionuclide distribution within the patient.

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Inside the patient, the gamma-ray photons originating from the radionuclide will emanate in all directions. Some of these photons will also interact inside the patient and be scattered at different energies and angles from the original radionuclide. For nuclear medicine studies, a collimator will be attached to the front of the gamma-camera head to produce a projection image. In its simplest form, a collimator consists of a large number of holes drilled in a lead slab, and is used to block out all gamma rays that are heading towards the crystal except those that are traveling at right angles to the plane of the crystal. Radiation entering at an angle to the crystal is absorbed by the lead, and only those gamma rays entering along the axis of the holes pass through. Unlike a lens, as used in visible-light cameras, the collimator attenuates most (>99 %) of incident photons and thus greatly limits the sensitivity of the camera system. Relatively large amounts of radiation must be present to provide enough counts to form an image. Most gamma cameras have several different collimators that are matched to the particular examination (Anigstein et al., 2005). For each collimator, the diameter and depth of each hole varies as does the thickness of lead between each hole (septum thickness). Each collimator is designed to provide a specific balance between the amount of gamma radiation allowed through (sensitivity) and the sharpness of the images (the spatial resolution). Collimators decrease the sensitivity of the detector system, as they restrict the number of photons that can reach the crystal. Removal of a collimator greatly improves the sensitivity of the camera because it allows more photons from a source to reach the detector. However, this also increases the background count rate recorded by the system. Since the minimum detectable activity is a function of counting efficiency and background counting rate, the use of or removal of collimators will affect the minimum detectable activity. The use of nuclear medicine cameras to screen patients following exposure to radionuclide contamination from a radiological dispersal device (RDD) or an improvised nuclear device (IND) incident requires advance planning, training and rehearsal; it cannot be implemented ad hoc in an emergency. 8.3.2.5 Pulse-Height Analyzer and Radionuclide Windows. A nuclear medicine gamma camera incorporates, as part of its design, a pulse-height analyzer that allows only photons within a predetermined energy range to be counted by the camera system and used

108 / 8. RAPID DETERMINATION OF INTERNAL CONTAMINATION to form the image. The operator is able to use the pulse-height analyzer to select energy ranges. The chosen energy range is typically centered on the primary photopeak(s) in the energy spectrum of the radionuclide of interest, and the range is usually set to encompass 10 % of the photopeak energy above and 10 % below the photopeak. Adjusting the range or width allows photons to be counted only in the photopeak region, or if desired, over much or all of the entire gamma-energy spectrum. For example, 18F, a common nuclear medicine radionuclide, produces a photon with photopeak energy of 511 keV. It is normally found as a “factory preset radionuclide” in the cameras’ selection menus. The width is found to be listed as 20 %, and therefore the energy range that the pulse-height analyzer will allow to be counted is 460 to 562 keV. The range is sometimes referred to as the “energy window.” Radionuclide windows that are not preset can be programmed into a camera’s computer. The nuclear medicine technologist has access to the predefined radionuclide list in the setup menu for each patient diagnostic acquisition. It is usually possible for the technologist to alter the photopeak energy or width for factorydefined radionuclides already programmed into the computer, if necessary. It is sometimes possible for the technologist to enter custom (i.e., user-defined radionuclides) that do not appear on the factory-installed list, and save these as new radionuclide windows. For example, 137Cs (662 keV photopeak) may be found as a predefined choice on some cameras, but not on others. Temporary adjustments to the photopeak energy and window width may be possible if the technologist is sufficiently familiar with the camera system. In general, it can be assumed that the radionuclides of concern in this Report will not be found on the standard list of factory-installed radionuclides on the gamma cameras. Predefined radionuclides windows can usually be adjusted to encompass the gamma spectrum for these radionuclides. Rather than define a new radionuclide window as a custom addition, it may be easier to use radionuclides already present in the predefined list and simply expand the acquisition width to encompass the photopeaks of the radionuclides of concern to this Report. Table 8.2 lists several commonly defined radionuclides on these cameras and their predefined widths, as well as the radionuclides of concern and settings required to count them. If the use of a nuclear medicine camera is to be part of a facility’s emergency plan and windows will be predefined, it is recommended that the camera manufacturer’s service personnel or a medical physicist experienced in nuclear medicine be involved in selecting the correct window settings. Setting up an emergency selection sequence in advance would be ideal and is encouraged.

Co

57

Co

Am

241

Ir

192

Cs

137

60

Other

Tl

201

Xe

133

I

131

Tc

99m

F

18

Clinical

Radionuclide

No

No

Sometimes

No

Yes

Yes

Yes

Yes

Yes

Yes

Preprogram Medical Setting

75 (40)

81 (20)

364 (20)

140 (15)

122 (15)

511 (20)

Predefined Energy Width (keV, % width)

60

300 – 605 (various)

662

1,173, 1,333

Photon Energy (keV)

Xe

F

F

133

18

18

Highest available energy

Alternate Predefined Radionuclide

100

100

100

200

Adjust Width (%)

TABLE 8. 2—Alternate settings for a nuclear medicine camera for possible radionuclides of concern (Anigstein et al., 2005). 8.3 DIRECT SCREENING FOR INTERNAL CONTAMINATION

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110 / 8. RAPID DETERMINATION OF INTERNAL CONTAMINATION Some radionuclides of concern, including 90Sr, 235U, 238U, 238Pu, Pu, and 241Am cannot be suitably detected by a gamma camera, especially at the level of 1 CDG.

239

8.3.3

Measuring Internal Contamination with Hospital Equipment

The choice of instrument for screening patients who may be internally contaminated depends on utility of the instrument as well as capability of the instrument to detect the radionuclide. A comparison of the suitability of various types of equipment for radionuclides of interest is shown in Table 8.3. 8.3.3.1 Using GM Survey Meters to Assess Internal Contamination with Certain Gamma-Emitting Radionuclides. The most likely instrument available to quickly setup a screening station is the GM survey meter. A portal monitor that incorporates GM tubes could also be used for this purpose, but setup time may preclude its immediate use. Preplanning would need to include calibration of the portal monitor. For GMs, count rates corresponding to 1 CDG for selected gamma-emitting radionuclides are provided in Section 8.5. This screening method assumes that the patient is free of external contamination. If a reading from the chest [antero-posterior (AP)] indicates that a patient has an intake in excess of the CDG, it would be prudent to measure from the back [postero-anterior (PA)] position to confirm that the indicated activity is from an intake rather than external contamination. 8.3.3.2 Using a Gamma Camera or Thyroid-Uptake Probe to Assess Internal Contamination with Certain Gamma-Emitting Radionuclides. A gamma camera or thyroid-uptake probe can be used to detect internal contamination as low as ~37 kBq (1 μCi) of a medium-energy, gamma-emitting radionuclide if the collimators are not used on the cameras. The collimators may have to be removed because only the highest energy radionuclides are visible to the detector with collimation in place. The collimators also provide protection to the crystal assembly, so care must be taken not to damage the crystal. Using a gamma camera without the collimator in place during an emergency situation could lead to detector damage and loss of warranty. The camera energy window should be set as wide open as possible around the principal photopeak of the contaminant radionuclide. Once the camera is setup, it should be calibrated with a phantom (commercial phantom or carboy) loaded with water that has been spiked with a known quantity of a radionuclide identical to the internal contaminant or with energies similar to the internal

Co (M)

Pu (M) 0.0048

0.0041

0.056

0.056

0.056

29

34

0.2

4.1

17

(0.13)

(0.11)

(1.5)

(1.5)

(1.5)

(770)

(910)

(5.5)

(110)

(460)

MBq (μCi) in Whole-Body at 24 h at 1 CDGb

Poor

Poor

Poor

Poor

Poor

Good

Good

Good

Poor

Good

GM

Poor

Poor

Poor

Poor

Poor

Excellent

Excellent

Good

Fair

Excellent

Portal Monitorc

Poor

Poor

Poor

Poor

Poor

Excellent

Excellent

Good

Fair

Excellent

ThyroidUptake Probec

Poor

Poor

Poor

Poor

Poor

Excellent

Excellent

Good

Poor

Good

Gamma Camerac,d

a Indicates rate of absorption into blood from respiratory tract; M and F = moderate and fast rate of solubilization, respectively. V = vapor which would be solubilized very rapidly (ICRP, 1994a) b NCRP (2008a). c Based on detector efficiencies in Anigstein et al. (2005). d Most cameras are set to detect gamma energies up to a maximum of 700 keV.

Am (M)

241

238,239

U (M)

234,235,238

Ra (M)

226

Po (M)

210

Ir (M)

192

Cs (F)

137

I (V)

131

Sr (F)

90

60

Radionuclide (type)a

TABLE 8.3—Detectability of 1 CDG for an inhaled radionuclide of interest using hospital radiation measuring equipment (Anigstein et al., 2005; NCRP 2008a). 8.3 DIRECT SCREENING FOR INTERNAL CONTAMINATION

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112 / 8. RAPID DETERMINATION OF INTERNAL CONTAMINATION contaminant. Like the GM screening method, this screening method assumes that the patient is free of external contamination. If a reading from the chest (AP) count indicates that a patient has an intake in excess of the CDG, it would be prudent to conduct a count from the back (PA) position to confirm that the indicated activity is from an intake rather than external contamination. The two counts should be averaged to determine whether the CDG has been exceeded. 8.3.4

Wound Monitoring

Blood smears of a wound may be suitable as a qualitative measure to identify initial wound contamination and to make some crude screening decisions for therapy. For intermediate- and highenergy beta-gamma emitters, lack of detection of radioactive material by direct counting of a wound and associated blood smear would suggest that dose intervention therapy is not indicated. Alpha emitters can be masked by the wet environment of a wound. 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 or alpha-scintillation probe). Radioactive blood smears are an indication of wound contamination and may indicate potential systemic uptake. Wound surveys to estimate the quantity of radioactive contamination in a wound can be accomplished for high-energy photon or beta emitters using portable GM instruments. Such surveys are likely to be most useful to monitor the progress of decontamination efforts. Once decontamination efforts are complete, direct measurement using simple or sophisticated gamma-ray spectrometry systems is preferred to quantify residual activity. Such measurements ideally should be coupled to direct and indirect bioassay methods appropriate for the radionuclide(s) of interest. Low-energy photon emitters (e.g., 241Am) and alpha emitters (e.g., 238, 239Pu) are substantially more difficult to monitor in a wound setting because attenuation can occur if radioactive material is embedded in tissue. Indirect bioassay is an important follow-up consideration for such circumstances. Measurements from blood smears (or skin breaks) >1,000 dpm transuranic alpha activity or 200,000 dpm beta-gamma activity, or indications of transuranic alpha wound activity of nominally 400 Bq (10 nCi) suggest that dose intervention therapy might be warranted. As noted by Carbaugh (2007), the choice of any single number for such indications of therapy is highly subjective. The numbers offered here are considered as general guidance, not rigorous quantitative values. Broader coverage of the entire topic of radionuclides in wounds is provided in NCRP (2006).

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8.4 Indirect (in vitro) Determination of Internal Contamination Indirect (in vitro) measurement of internally-deposited radionuclides involves analyzing material removed or excreted from the body. Typical media for measurement include nasal swabs, urine samples, and fecal samples. Blood samples (e.g., residual blood from clinical drawings, or drawings specifically for radiochemical analysis) are not used routinely for internal contamination assessment. Nasal swabs can be used as an indicator for the possibility of internal contamination, whereas fecal samples provide direct evidence that internal contamination has occurred, and urine samples are particularly valuable to demonstrate systemic uptake. Fecal and urine samples are typically part of a radiobioassay program to assess intake and dose (IAEA, 2000; NCRP, 1987). Indirect bioassay samples require analysis by a properly equipped and staffed radiochemistry laboratory, although very limited screening of samples using portable survey instruments may be possible for some radionuclides. Prompt radiochemical laboratory analysis of a large population sample collection effort will be challenging, having to deal with issues of sample collection, packaging, shipping, analysis and reporting of results. 8.4.1

Nasal Swabs

Nasal swabs are crude indicators of intake and should be considered qualitative and not quantitative with regard to dose intervention therapy decisions. Nasal swabs may be collected by taking a gentle smear of each nare using a cotton tipped applicator or a small 5 cm (2 inch) gauze pad (a 47 mm air filter paper is often used in occupational exposure settings) wrapped around an applicator. The use of the wrapped gauze pad or air filter facilitates detection by allowing a flat geometry to be used for subsequent counting with a radiation detector. While not a reliable indicator, the presence of radioactive material on nasal swabs may indicate an inhalation exposure. Experience has shown that if one nare shows relatively-high contamination whereas the other shows little or no contamination (e.g., an order-ofmagnitude difference between them), the subject may have brushed his/her nose with a contaminated hand or finger, rather than actually inhaled contamination. If the nasal swabs show similar contamination in both nares, and there is clearly identified facial contamination, particularly in the mouth or nose region, then inhalation is probable. Guilmette et al. (2007) analyzed thousands of nasal swabs and noted that the nasal swab measurement from one nostril was typically greater than from the other nostril. This was

114 / 8. RAPID DETERMINATION OF INTERNAL CONTAMINATION attributed to the nasal cycle, which is common to most people, and which results in preferential airflow through one nostril. They suggest either counting the samples together or summing the activities from both sides as a more reliable measurement and indicator of inhalation intake. The absence of radioactive material cannot definitively rule out an inhalation exposure. If the subject is a mouth-breather due to allergies or sinus congestion, the nasal swabs may not be representative. Additionally, if the subject sneezes or clears his or her nasal passages prior to counting, the subsequent count may not represent actual intake. The extrathoracic region of the respiratory tract is also subject to very rapid clearance (ICRP, 1994a), so if swabs are obtained an hour or more after an exposure incident, normal clearance may have already have effectively eliminated nasal contamination. However, studies by Smith (2003) indicate that there is a second retention component of particles in the nasal airways that make nasal swabbing beyond 1 h still useful. Because of the highly variable and rapidly changing nature of nasal retention, results of nasal swabs of the public should be used with caution and should not be used as a sole decision for dose intervention therapy. Guidance developed for occupational exposures at the DOE Hanford Site (Carbaugh, 2007) suggests that therapy may be warranted for inhalation of transuranics (e.g., plutonium, 241 Am), based on nasal swabs in excess of 20 Bq (1,000 dpm) alpha in each nare. Therapy for 90Sr or 137Cs inhalation might be indicated based on nasal swabs in excess of 1,700 Bq (100,000 dpm) beta in each nare. 8.4.2

Urine Samples

Section 7.2.2 provides advice on collection of single-void urine samples for radionuclide analyses. This section and Appendix G provide a general discussion of collection and analysis of urine samples. The preferred method for rapid screening by indirect bioassay in an emergency or disaster situation is collection of a single void (or nominally a 50 to 100 mL sample) of urine. Samples taken earlier than 4 to 6 h post-exposure require careful interpretation because metabolism and transfer to urine may be insufficient to provide a valid indication of intake. Collection of urine is relatively straightforward, convenient, and easy to manage. The patient should be given the standard hospital instructions for clean-catch procedures to prevent or minimize external or cross-contamination of the sample during collection and handling. Urine samples should be collected in clean, tightly sealed containers. In the absence of standard sample containers, any clean

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plastic or glass liquid-tight container could be used. Samples must be labeled to identify the person and a sample collection date. The sample collection time should be noted if the sample is collected within 4 to 8 h of the exposure incident. Most nuclear medicine departments possess a NaI(Tl) well counter that can be used to assay urine bioassay samples. The well counter is ideal for assaying urine for gamma-emitting radionuclides such as 60Co, 137Cs, or 192Ir, although inhalation of insoluble forms of these radionuclides would not be expected to result in a significant amount of radioactive material in urine. The well counter would need to be calibrated with the radionuclide of interest, and the calibration source would need to approximate in shape and volume the vial or tube used to contain the urine. Single-void urine samples may be suitable for use as a screening tool for identifying individuals who may exceed the CDG or who should be part of a more rigorous bioassay and dose-assessment protocol. Generally, they are not good for doing definitive dose estimates requiring some kind of normalization (with accompanying uncertainties) to a total 24 h excretion. Total 24 h (or approximate 24 h) urine samples are the preferred tool for measuring systemic excretion for dose assessment purposes. Such dose assessments are addressed in Section 11 as part of the long-term follow-up activities discussed in this Report. Appendix G provides guidance for the preparation of biological samples for shipping. 8.4.3

Blood Samples

Blood samples are not likely to be an effective tool for radionuclide analysis as part of screening for a radionuclide exposure. The concentration of a radionuclide 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. Sun (1997) suggested the suitability of radiochemical analysis of blood samples for plutonium dosimetry but the highly sensitive analytical method described is not suitable for rapid screening nor for large numbers of people. If high-level exposures from an external radiation source are suspected, blood samples are indicated for differential complete blood cell counts (in particular circulating lymphocytes) as a baseline for clinical monitoring. Blood samples are also suitable for cytogenetic dosimetry evaluation of a relatively-high dose (nominally 250 mGy) from external gamma or neutron radiation exposures. Such samples do not require any special considerations in collection beyond normal clinical practice. They do not pose any radiological hazard to attending personnel.

116 / 8. RAPID DETERMINATION OF INTERNAL CONTAMINATION 8.4.4

Fecal Samples

The logistics associated with the collection and analysis of fecal samples renders them relatively ineffective as rapid screening tools, especially when dealing with a large group of potentiallyexposed people. Fecal samples for radioanalysis supporting dose assessments need to be total voidings, not stool smears or swabs. They could be a very useful tool for individual cases of inhalation of a radionuclides in a relatively-insoluble form in post-crisis dosimetry investigations or a long-term registry program, but are not pertinent to initial, emergency screening. Fecal samples should be collected in clean, airtight, sealable plastic containers (e.g., commercial fecal sample kit or containers similar to one pint to two quart ice cream containers). Samples must be labeled to identify the person and sample collection date. 8.4.5

Analysis of Indirect Bioassay Samples

The analysis of indirect bioassay samples by radiochemical means requires analytical radiochemical procedures and equipment not typically found in clinical laboratories. A relatively small number of commercial and government radiochemistry laboratories provide support to the nuclear industry for these analyses, and processing times are such that they probably would not be considered responsive for rapid screening of populations. The large numbers of samples that might be anticipated following a large scale incident would probably create a long-term processing backlog at these facilities. CDC can provide assistance with regard to some types of radiochemical analyses.5 Rapid screening of indirect bioassay samples using survey instruments could provide some information but if levels are sufficiently high to be detectable using survey instruments, direct detection by survey of the individual is likely to be of more value to early response and screening. For some of the radionuclides of concern to this Report, rapid screening of bioassay samples is simply not possible. For those radionuclides, decisions regarding triage and therapy must be made on the basis of other information, such as patient history. Efforts are underway at several laboratories to develop and/or improve radioanalytical methods and capacity for analysis of urine samples. One example is the work being conducted by the Inorganic and Radiation Analytical Toxicology Branch of CDC (2005c). 5Jones,

R.L. (2007). Personal communication (Centers for Disease Control and Prevention, Atlanta, Georgia).

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CDC has been working with other researchers in the development and refinement of methods for quick assessment of internal contamination using small (spot) urine samples (<100 mL). CDC is able to use these methods to quickly analyze for 232Th, 235U, 238U, and 239Pu and research is ongoing to expand the list of radionuclides available for this type of analysis. CDC is also in the process of obtaining Clinical Laboratory Improvement Amendments approval for a gross-alpha/gross-beta method using liquid scintillation.6 In addition, there is work under way to use gamma spectroscopy to conduct quick screening of urine samples to expedite analysis of those that require detailed isotopic analysis. CDC and other laboratories are expanding the use of inductively coupled plasma mass-spectrometry methods in the detection of low levels of radioactive contamination that can be used for a large number of samples. These methods have been successfully used to measure both toxic and essential elements in biological matrices such as urine in as little as 1 mL (Caldwell et al., 2005; Pappas et al., 2004). These methods have short counting times and require less processing than those analyzed by traditional radiochemical methods, thus allowing for the quick screening of possibly thousands of low-level samples per day. A method for alpha spectroscopy of evaporated water residues developed by the New York State Department of Health Wadsworth Center consists of evaporation of drinking water, flaming of the planchets, and alpha-spectroscopic measurements using a grid ionization chamber. The method can identify and quantify radioactivity concentrations ≥3 mBq L–1 in a matter of several hours, whereas determination of sub mBq L–1 levels is achievable in 1 d. The method was successfully used to measure 230Th, 238U, 239Pu, 241 Am, and 244Cm, and it provides a fast way to identify or screen urine samples during a radiological emergency involving alpha radioactivity (Semkow et al., 2009). 8.5 Rapid Screening of Persons to Identify Radionuclide Intakes that Exceed the Clinical Decision Guide Hurtado (2006) and Juneja (2011) studied the response of different types of survey instruments to emissions from internallydeposited radionuclides using two hybrid computational phantoms to model an adult male and an adult female who had been internally contaminated by an acute inhalation or ingestion intake of 6Jones,

R.L. (2007). Personal communication (Centers for Disease Control and Prevention, Atlanta, Georgia).

118 / 8. RAPID DETERMINATION OF INTERNAL CONTAMINATION 241

Am, 60Co, 137Cs, 131I, or 192Ir. For each of four different detector types (including GM) and various screening distances, they estimated the net counts per minute corresponding to committed effective doses of 50, 250, and 500 mSv using Monte-Carlo radiation transport simulation. The effective dose threshold for the CDG based on stochastic effects is 250 mSv for an adult and 50 mSv for a child. The resulting count- rate values are provided within a series of handbooks available at the CDC website (CDC, 2009). As discussed earlier in this section, the radiation detection instrument that hospitals and local emergency responders are most likely to possess is a GM survey meter. In this Report, tables were selected from CDC (2009) that provide externally-detectable count rates in a GM survey meter for internal contamination of adult male and female subjects resulting from an inhalation exposure to an aerosol with an aerodynamic particle size of 1 μm AMAD and a Type M lung absorption rate for 60Co and 192Ir, and Type F lung absorption rate for 137Cs. Americium-241 was not included in this rapid screening method due to the difficulty of detecting 241Am. If 241Am is identified as the radionuclide in the incident, indirect (in vitro) measurement is recommended to determine whether the CDG has been exceeded. Also, 131I was not included because of the low CDG for 131I. If 131I is identified as an important radionuclide in an incident and a GM measurement over the thyroid indicates that internal contamination is present, KI should be administered as soon as possible per FDA (2001) recommendations (Section 7.6). At the time this Report was being completed, tables of calculated net count rates corresponding to different effective doses were available for an adult male and an adult female but not for children (Hurtado, 2006; Juneja, 2011). Recognizing that a population exposed to a radiological or nuclear incident will likely include children, this Report contains additional tables of estimated values for a child. A representative age of 10 y was chosen arbitrarily for the child because it is approximately the middle of the age range for children. Because the CDG based on stochastic effects in a child is defined as the amount of internal contamination that would result in an effective dose of 50 mSv, the count rates for the CDG values associated with 60Co, 137Cs, and 192Ir in a child were estimated by adjusting the count rates in the 50 mSv tables for an adult female (CDC, 2009). The count rates were adjusted for physical geometry and radiation attenuation and also for biokinetics to account for differences in body mass and metabolism between an adult and a child. For a given internal body burden, Bolch and Juneja7 estimated that the count rate from a 10 y old child would be ~1.2 times higher than

8.5 RAPID SCREENING OF PERSONS TO IDENTIFY RADIONUCLIDE

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from an adult female due to geometry and attenuation differences between an adult female and a 10 y old child. Ansari8 computed a similar ratio. The ratios of effective dose per unit activity in an adult female to that in a 10 y old child for these three radionuclides are 0.67 for 60Co, 0.97 for 137Cs, and 0.68 for 192Ir, respectively (NCRP, 2008a). Thus, count rates from a 10 y old child were estimated from the product of the count rate for an adult female (50 mSv), a geometry and attenuation ratio of 1.2, and a radionuclide-specific effective dose ratio for these two age categories. While this method is somewhat crude, it is judged to provide adequate guidance until more rigorous determinations can be made for various ages of children. Emergency planners should periodically consult the CDC (2009) website for updates on all tables, which may include updates on the radionuclides used in this Report as well as other radionuclides, ages and distances. Table 8.4 provides GM count rates for 1 CDG of 60Co, 137Cs, or 192Ir at 6 cm and 30 cm directly in front of the sternum (AP) or directly in the middle of the back (PA) 1 h after exposure to each radionuclide. Count rates for the adult male and female were obtained from the tables published on the CDC website (CDC, 2009) and rounded to two significant figures. Count rates for a 10 y old child were calculated as explained above. Additional tables that provide count rates calculated in the same manner for various times from 1 to 72 h post-exposure are provided in Appendix F. CDC continues to refine and expand their tables. Emergency planners should consult the CDC website (CDC, 2009) periodically and should incorporate updates into their local emergency plans.

7Bolch, W. and Juneja, B. (2010) Personal communication (University of Florida, Gainesville, Florida). 8Ansari, A. (2010) Personal communication (Centers for Disease Control and Prevention, Atlanta, Georgia).

10 y old child

Female

Male

Subject

Cs Ir

137 192

Co

Ir

192 60

Cs

137

Co

Ir

192 60

Cs

Co

137

60

Radionuclide

2,900

2,300

2,500

17,800

10,000

15,500

14,400

6,300

12,900

AP

6 cm

1,100

730

870

10,200

5,000

9,600

9,500

3,800

9,200

PA

1,700

1,200

1,600

6,600

3,200

5,400

6,000

2,300

4,800

AP

30 cm

Distance from Sternum or Middle of Spine

560

400

520

3,400

1,700

3,200

3,300

1,500

3,100

PA

TABLE 8.4—Adult male, adult female, and 10 y old child net count rate (counts per minute) corresponding to 1 CDG for various radionuclides at 1 h after exposure using a GM survey meter (CDC, 2009).

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9. Medical Management of Internally-Contaminated Persons 9.1 Introduction At the scene of an incident where people may have been contaminated with radionuclides, emergency responders may be called upon to provide first aid or even resuscitation. The highest priority is treatment of conventional injuries that may be life threatening (e.g., bleeding, shock). Similarly, after transported patients reach the emergency room, trained medical personnel should evaluate their medical condition and ensure that all patients are medically stabilized prior to external decontamination and screening for internal contamination. Most physicians are unlikely to have had experience managing radiation injuries or patients who are internally contaminated. Additionally, the majority of clinical experience with medical management of persons having internal radionuclide contamination has been limited primarily to industrial accidents involving a few individuals in settings where industrial health-physics protocols and sufficient resources were present. Projected scenarios involving RDD incidents and other radiological terrorism incidents, however, differ in significant ways from historical radiation accidents. The geographical dispersal of a radionuclide(s) could cover an area ranging in size from a few city blocks to several square kilometers, with hundreds to thousands of people having external and internal contamination, but with little external radiation exposure (DHHS, 2009). There could also be a delay in the identification of the specific radionuclide(s) utilized leading to a delay in evaluation or treatment of patients, since countermeasures vary dependent on the radionuclide. In the case of a nuclear detonation, fallout will reach across many jurisdictions, potentially involving multiple states (HSC, 2009). Resources and availability of experienced medical and health-physics personnel may also be limited during early response activities, which will contribute additional challenges to healthcare providers making diagnostic and treatment decisions. Medical personnel should be aware that anyone who is internally contaminated also may have received a large external dose of 121

122 / 9. MANAGEMENT OF INTERNALLY-CONTAMINATED PERSONS radiation, especially those who were located near to an RDD or those who were contaminated as a result of an IND incident (Hrdina et al., 2009). Subsequent to the emergency response, dose reconstruction based on plume modeling or environmental sampling may be performed to estimate the radiation dose to members of the public who were exposed during the incident, but dose estimates are highly unlikely to be available to treating physicians in a timely manner nor would they be available for any one specific patient based on these methods. Biodosimetry, the process of estimating a dose to a patient using laboratory or other clinical testing, should be conducted on some patients, particularly those who were close enough to the device to have been exposed to a potentially high dose of radiation (>1 Gy). The most common method of biodosimetry for suspected high-dose exposures utilizes serial complete blood counts for evidence of lymphocyte depletion. Subsequently, 6 to 24 h after the incident, a blood sample could be taken for cytogenetic analysis, also known as dicentric chromosome assay (DHHS, 2009). Cytogenetic analysis can estimate a radiation dose due to whole- or significant-body irradiation with a threshold of 0.05 Gy and possibly even lower, but the results may take up to a week before they are available (DHHS, 2009). This should be coordinated with AFRRI (Prasanna et al., 2005) or the Radiation Emergency Assistance Center/Training Site (REAC/TS, 2010) who will arrange for the analysis as part of the coordinated federal response to the incident. Unfortunately for the treating physician, lymphocyte depletion and cytogenetic analysis are not generally suitable for assessing partial-body exposures or internally-deposited radionuclides (DHHS, 2009). Evaluation and medical management of ARS are beyond the scope of this section and readers are referred to other textbooks and articles related to this topic (e.g., Hrdina et al., 2009). Of special note is the Radiation Event Medical Management System that was developed to provide physicians with a just-in-time, algorithmbased set of guidelines and a comprehensive tool to assist in the evaluation and treatment of radiation patients (DHHS, 2009). 9.2 General Clinical Guidance for the Treatment of Internal Contamination Physicians responsible for the treatment of potentially-exposed individuals must weigh the relative risks of short- and long-term biological effects. The risks associated with radionuclide intakes are largely long-term in nature, also called stochastic effects, and depend not only on the type and concentration of the radionuclide absorbed, but also on the health and age of the exposed individual. The potential for development of cancers of the hematopoetic system,

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lung, liver, thyroid, bone, among others, are the primary long-term health concerns, as are fibrotic changes in tissues such as lung, which may lead to restrictive lung disease and other chronic debilitating conditions. The method of reducing these risks is removal of the radionuclide(s) from the body. The long latency of these conditions means that evaluation and treatment of internal contamination should not take precedence over treatment of conventional injuries that may be acutely life-threatening. However, early recognition of internal contamination provides the greatest opportunity for radionuclide removal (AFRRI, 2003; Gerber and Thomas, 1992; NCRP, 1980; 2008a). The presence of internal contamination is rarely lifethreatening and decorporation should not take precedence over treatment of conventional injuries that may be acutely life threatening. For the purposes of this section, the term decorporation agents refers to drugs that increase the rate of elimination or excretion of absorbed, inhaled or ingested radionuclides. Decorporation can be accomplished in several ways: by reducing absorption, preventing incorporation (uptake of radionuclides within organs), and by promoting elimination or excretion of absorbed radionuclides. The effectiveness of most decorporation agents for the treatment of internal radionuclide contamination has not been tested in humans because the occurrence of large accidental or nonaccidental radionuclide intakes is rare (FDA, 2006a). Due to renewed interest in the possible need to use decorporation agents after an RDD or IND, investigations of new approaches to radionuclide decorporation are underway. Recent reports of progress in this area are available in the Proceedings of the 10th International Conference on Health Effects of Incorporated Radionuclides (Guilmette, 2010a; 2010b). The following guidelines should have high importance in planning for the medical management of internal contamination in response to a mass casualty radiological incident: • Radionuclide identification is necessary prior to any bioassay analyses or use of decorporation agents. • During the early phase of a large-scale incident, it is likely that significant radiation-safety support, instrumentation, and bioassay support will not be immediately available. • Initial treatment decisions for radionuclide intakes may therefore be based on the medical history, the magnitude of external radionuclide contamination on the patient, and the patient’s proximity to the incident.

124 / 9. MANAGEMENT OF INTERNALLY-CONTAMINATED PERSONS • External dose is expected to be small in an RDD incident except for those possibly near the device or for cases involving embedded shrapnel (Musolino and Harper, 2006). • Internal radionuclide contamination by itself is most likely to have an asymptomatic presentation. Symptoms indicative of the ARS such as nausea and vomiting will be rare, and few, if any, patients exposed to radionuclides from an RDD will be contaminated to a level that requires ARS treatment (DHHS, 2009). However, if vomiting does occur, time-toonset of vomiting combined with lymphocyte depletion kinetics, can be used to estimate radiation dose. Absence of the signs and symptoms of ARS does not eliminate the need for further evaluation for internal radionuclide contamination if history and corroborating evidence indicate possible internal contamination. 9.3 Making Requests for Radiological Countermeasures If it is determined that decorporation agents are required for the treatment of internally-deposited radionuclides, physicians and hospitals may need to contact local or state public-health agencies to obtain the necessary treatments since many of them are not common in hospital or commercial pharmacies. The local public-health authority would be responsible for making the request for countermeasures through the usual chain of command to the state emergency management who would forward the request to CDC who would release the countermeasures from the SNS (CDC, 2008). The SNS consists of equipment, supplies and pharmaceuticals that might be needed in the event of a large-scale disaster or catastrophe. Examples of equipment and pharmaceuticals in the SNS are vaccines, burn-supplies, and antibiotics. The SNS is intended as an all-hazard stockpile but includes drugs for use in radiological or nuclear incidents. The stockpiled radiological medical countermeasures include the following pharmaceuticals in the SNS vendormanaged inventory: Prussian blue capsules [Radiogardase® (Heyltex Corporation, Katy, Texas)]; calcium- (Ca) and zinc- (Zn) DTPA ampoules for intravenous (IV) or inhalation administration; filgastrim [Neupogen® (Amgen, Inc., Thousand Oaks, California)]; and KI. Decorporation agents needed to treat for internally-deposited radionuclides may be available nearby as a result of a SNS forwardplacement program with local and state public-health and radiation-control programs. This program has placed population-based caches of Ca- and Zn-DTPA in the hands of local and state authorities to facilitate more timely administration of these countermeasures.

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Additionally, some municipalities and states may have purchased their own supplies of countermeasures to provide initial treatment until the SNS vendor-managed inventory arrives, and local hospitals or pharmacies may stock KI as a blocking agent or Neupogen® to treat patients undergoing radiotherapy or chemotherapy. Other managed-inventory assets in the SNS, such as ventilators or nonpharmaceutical consumables may also be requested. These assets are shipped within 4 to 12 h to state prearranged locations. SNS asset custody is transferred from the federal government to state possession, and then from state possession to local municipalities at prearranged dispensing sites that have been deemed eligible to manage the assets. Local distribution and medical management recommendations for use of the SNS managed-inventory pharmaceuticals become the responsibility of the local public-health authority. For this reason, it is imperative that local municipalities and public-health authorities plan for the requesting and distribution of SNS assets. Local municipalities and public-health agencies must develop procedures for requesting, receiving and distributing Strategic National Stockpile (SNS) assets. This planning should include providing the diagnostic and medical management guidelines to healthcare providers for the use of decorporation agents for internally-deposited radionuclides. Lists of possible decorporation agents for radionuclides beyond those covered in this Report are given in NCRP Report No. 161 (NCRP, 2008a). Localities that rely on the SNS for radiological medical countermeasures will be faced with a delay in delivery of these medications to the point-of-care which will delay treatment of patients identified with high levels of internal radionuclide contamination. It may be impractical to keep patients in emergency departments or radiological screening stations for long periods of time in order to await arrival of medications. These patients will require clear discharge instructions on how and where to obtain the necessary medications. Local public-health authorities should have plans in place for such an eventuality. 9.4 Medical Management Decisions for Potentially-Contaminated Persons 9.4.1

Triage

Standard triage as practiced by emergency departments assumes access to adequate resources for all patients. Disaster triage has as

126 / 9. MANAGEMENT OF INTERNALLY-CONTAMINATED PERSONS its goal doing the greatest good for the greatest number of people based on the limited resources available. Concepts of radiological triage may be applied to either concept of triage. Radiological triage determines which mode(s) of radiation exposure a victim could have incurred: • • • •

irradiation from an external source; internal radionuclide deposition; external radionuclide contamination; or embedded source shrapnel.

The mode(s) of radiation exposure for a given patient will dictate the types of radiological medical countermeasures and treatment of the patient. For example, a patient with a single mode of exposure to radiation from an external source would be potentially at risk for only ARS or local radiation injury, and therefore would not require any treatment for internal radionuclide deposition. However, a patient with an embedded source could potentially be exposed to high doses of radiation and internal radionuclide deposition requiring countermeasures that address ARS and internal contamination. The combination of significant conventional trauma requiring surgery (or burn care) and significant radiation exposure causing deterministic effects is known as combined injury. Combined injury, where it occurs, may result in increased morbidity and mortality, and should be a triage consideration (DHHS, 2009). Patients involved in an explosive radiological incident, such as an RDD may present to hospitals with both combined injuries and combined modes of radiation exposure (irradiation from an external source, internal contamination, external contamination, and embedded source shrapnel). Traditional trauma and medical care should take precedence over radiological concerns. 9.4.2

Prioritizing Children and Pregnant Women

Fetuses, infants, children and adolescents are more radiosensitive than the adult population to both deterministic and stochastic effects (AAP, 2003). Adults over 40 y of age are relatively unlikely to display stochastic effects due to the long latency period of these effects and to the protracted period of time over which dose is received from many internally-deposited radionuclides. The models presented by NA/NRC (2006) assess health risks from exposure to low levels of ionizing radiation show that lifetime attributable risk decreases with increasing age at exposure. The increased radiosensitivity for children and fetuses is reflected in a lower CDG for children and pregnant women. For this reason, physicians should prioritize children and pregnant women for both screening

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and treatments, especially if faced with large numbers of potentially internally-contaminated persons. 9.4.3

Medical Decisions During the Early Phases of a Radiation Incident

Treatment decisions for internal contamination should be based on the possible risk to the patient for both deterministic and stochastic effects. Historically, treatment decisions have been made empirically based on the patient’s history and presentation. Accurate quantitative evaluation of internal radionuclide contamination levels traditionally takes days to weeks, well beyond the time that most countermeasures are maximally effective. The CDG first described in NCRP Report No. 161 (NCRP, 2008a) and discussed in Section 7 denotes a radionuclide intake that indicates an action may be required by the treating physician such as more intensive screening or initiation of treatment. The CDG for radionuclides other than isotopes of iodine 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 of age for intake by adults and to 70 y of age for intake by children and pregnant women, that is in the range associated with guidance on dose limits for emergency situations (DOE, 2009b; FEMA, 2008a; 2008b; ICRP, 1991; 2005); 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 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 as discussed in Section 8. The CDG is applicable to members of the public and to workers. The prevention of deterministic effects, which could occur at levels above the CDG, should have the highest priority for decorporation therapy, if such treatments exist. Historically, deterministic effects from internal radionuclide contamination have been extremely rare. There have been six reported cases of ARS caused solely by internal contamination, and therefore, the deterministic effects from internal contamination are unlikely to be seen (Harrison et al., 2007; IAEA, 1998; Mettler, 2001; NCRP, 1980). The greatest number of exposed persons may have some internal contamination

128 / 9. MANAGEMENT OF INTERNALLY-CONTAMINATED PERSONS that will not cause deterministic effects, but they may still be at risk for long-term stochastic effects and may theoretically benefit from treatment. Calculating the CDG requires health-physics support and additional information such as results of chest or whole-body counts or measurements of activity in excreta to determine the patient’s intake. This information may not be available during the early phase of a response if resources and personnel are limited, and physicians may be faced with the need to make decisions based entirely on history or simple observations, such as a history of proximity to the incident or obvious external dust or dirt on the individual’s body. Later, victims may be treated based on simple qualitative (yes or no) results from external radiation survey screening, especially if the radionuclide is a predominant gamma emitter. As more radiation instrumentation assets and radiation-safety personnel become available, decisions to treat may be made on the basis of CDGs derived from internal contamination screenings. 9.4.4

Choice of Decorporation Therapy for InternallyDeposited Radionuclides

Internal contamination by radionuclides is often considered by physicians as a type of poisoning. Poison-control centers and associated toxicologists may be a useful resource for the treating physician and public-health agencies, especially when multiple treatment regimens exist. It is paramount that toxicologists and responding staff at poison-control centers be familiar with, and have access to, the current recommendations for the treatment of internally-contaminated individuals (NCRP, 2008a). The Radiation Emergency Assistance Center/Training Site (REAC/TS) should be used as a resource for information and assistance when considering the treatment of internally-contaminated individuals (REAC/TS, 2010). FDA has approved only a few decorporation agents for the removal of radionuclides from the body. Due to the infrequency of incidents, clinical experience with the medical management of many of the radionuclides is extremely limited or nonexistent. Historically, treatment decisions have typically been made on a case-bycase basis in close consultation with experts in the medical management of radiation incidents. Many of these treatments involved “off-label” uses of FDA approved drugs. In a mass casualty incident involving tens to thousands of potentially internally-contaminated

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individuals, treating physicians may not have expert guidance readily available for each individual and will have to rely on available written recommendations. In the following section, treatment recommendations are based on expert consensus from NCRP (2008a) and will be identified as either FDA approved or off-label use. Many of the agents mentioned are thought to be safe and effective, and remarkably free of side effects. However, relatively few administrations of these agents have been used in actual practice. For example, only a few thousand doses of DTPA were distributed in the past. Also, there have been few documented administrations of insoluble Prussian blue. When moderately larger numbers of people have been treated with these agents, it is possible that a different profile of drug side effects may emerge and clinicians should be alert to this possibility. 9.4.5

Self-Treatment

Many of the treatments for an internally-deposited radionuclide involve generically available treatments or over-the-counter medications. With the exception of KI, used to block thyroidal uptake of 131I, there are no circumstances in which NCRP recommends selftreatment for radionuclide removal. Any potential benefits of selftreatment are outweighed by the potential harm caused by persons overly anxious to treat themselves or their children who may carry treatment to an extreme and unintentionally cause induced toxicity. Additionally, many treatments require laboratory testing to evaluate efficacy, which would not be available to self-treating individuals. 9.4.6

Contaminated Wound Management

In responding to an incident involving radionuclide contamination, physicians should consider patient wounds to be contaminated with radioactive material until proven otherwise. Normally, general decontamination will have been performed by emergency responders, and decontamination of wounds will be left to the discretion of medical personnel. When wounds are contaminated with a radionuclide, the physician must assume that uptake (internal deposition) has occurred. Appropriate action is based on the radionuclide involved and its activity level. It is important to initiate measures that prevent or minimize uptake of the radioactive material into body cells or tissues (Table 9.1) (NCRP, 2006; REAC/TS, 2010). When monitoring contaminated wounds or irrigation fluids, gamma radiation is easily detected while beta radiation may prove more difficult to detect. Alpha contamination will be especially difficult to measure due to the fact that blood and debris can prevent

130 / 9. MANAGEMENT OF INTERNALLY-CONTAMINATED PERSONS TABLE 9.1—Management of radionuclide-contaminated wounds (adapted from NCRP, 2006; REAC/TS, 2010). • Contaminated wounds are first draped, preferably with a waterproof material, to limit the spread of activity. • Gently irrigate wound with copious amounts saline or water [irrigation with DTPA or other countermeasures may be considered (Section 9.4.7.1)]. More than one irrigation is usually necessary. • The wound should be monitored after each irrigation. Contaminated drapes, dressings, etc., should be removed for accurate results. • Following repeated irrigations, if decontamination is successful, the wound is treated like any other wound. • If the preceding decontamination procedures are not successful, and the estimated contamination exceeds the CDG, conventional debridement of the wound must be considered. Excision of vital tissue from a wound solely to remove radioactive contamination should be performed only upon the advice of a physician expert in radiological emergencies (NCRP, 2006). Debrided or excised tissue should be retained for health-physics assessment. • Embedded radioactive particles, if visible, can be removed with forceps or by using a water-pik. Puncture wounds containing radioactive particles, especially in the fingers, can be decontaminated by using an “en bloc” full thickness skin biopsy using a punch biopsy instrument. • Once removed, any radioactively contaminated embedded particles, shrapnel, or debris should be appropriately shielded by placing in lead containers that are available in nuclear medicine departments. • After the wound has been decontaminated, it should be covered with a waterproof dressing. The area around the wound should be decontaminated as thoroughly as possible before suturing or other treatment. • Contaminated burns, chemical or thermal, are treated like any other burn and may be gently rinsed but not scrubbed. Contaminants will slough off with the burn eschar. However, dressings and bed linens can become contaminated and should be handled appropriately.

detection of alpha particles. Additionally, an alpha probe would be required, and this is not typically available in a hospital emergency department. 9.4.7

Using DTPA on Radionuclide-Contaminated Wounds

Irrigation of wounds with DTPA may be considered for some radionuclides such as the lanthanides and actinides (NCRP, 2006). 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

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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 inhalational administration of DTPA, but care must be taken to avoid over-dosing with DTPA because the amount absorbed by the wound tissues cannot be measured. Further discussion of decontaminating wounds appears in NCRP Report No. 161 (NCRP, 2008a). 9.5 Medical Management of an Americium-241 Intake 9.5.1

Overview

Americium is a member of the actinide series. Americium-241 has a radioactive half-life of ~432 y and emits alpha particles with a total energy of ~5.5 MeV per transformation and x and gamma rays. The retention of inhaled 241Am in the lungs and its absorption from the GI tract depend on the in vivo solubility of the form(s) inhaled or ingested. Americium-241 reaching the systemic circulation after absorption from the lung, GI tract, or a wound is deposited primarily in the skeleton and liver where it is avidly retained. Further information on the biokinetics and dosimetry of internallydeposited 241Am can be found in NCRP Report No. 161 (NCRP, 2008a) and ICRP Publication 67 (ICRP, 1993). 9.5.2

Treatment

Chelation with Ca- or Zn-DTPA given by IV or inhalation (FDA, 2004): Adults and Adolescents: Start: Ca-DTPA, 1 g in 5 mL slow IV push over 3 to 4 min first day (Table 9.2). 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, Ringer’s 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. 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.

Notes

Mechanism of action: heavy metal chelation.

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. After 24 h, obtain a 24 h urine bioassay to assess efficacy of the first dose and as a basis to consider further therapy.

Administer

First day administer Ca-DTPA, then for maintenance use Zn-DTPA. FDA approved alternate modes of 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 sterile water, Ringer’s lactate or normal saline infused over 30 min.

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

DTPA is not recommended for chelation with uranium.c

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

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

Contraindications

TABLE 9.2—Ca- or Zn-DTPA treatment (NCRP, 2008a).a

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

Occasional local skin reaction may occur.

Blood pressure during administration.

Monitor Clinically

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

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

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

b

FDA (2010a). Pregnancy categories are defined in Appendix J. c Henge-Napoli et al., (2000).

a

For children, <12 y of age, 14 mg kg–1 slow IV push over 3 – 4 min.

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134 / 9. MANAGEMENT OF INTERNALLY-CONTAMINATED PERSONS Administer one dose in mass casualty incidents. The first dose can be the most important for chelation of radionuclides inhaled in a relatively soluble form. After 24 h, obtain a 24 h urine bioassay to assess efficacy of the first dose and as a basis to consider further therapy. Traditionally, most patients have received only one dose IV. On the other hand, treatments for a few high-level exposure cases have continued for months or years. Toxicity is due to chelation of needed metals, such as zinc and manganese. Zn-DTPA should be used exclusively in pregnant patients. Efficacy of DTPA is greatest if administered within 24 h of intake, but can still be effective if given later (Marcus, 2004; NCRP, 2008a). Ca- and Zn-DTPA are FDA approved for the treatment of individuals with known or suspected internal contamination with plutonium, americium or curium (FDA, 2004). 9.5.3

Medical Follow-Up After Treatment

Repeated bioassay measurements should be obtained after treatment to evaluate the effectiveness of radionuclide removal. In some accident cases, therapy has continued for months or years. In cases such as these, follow-up may include radiochemical analyses of urine samples, supplemented by fecal samples, as needed. 9.6 Medical Management of a Cesium-137 Intake 9.6.1

Overview

Cesium, like potassium, rubidium, and several other elements is a member of the alkali metals family. The radionuclide of most interest here is 137Cs, which has a radioactive half-life of ~30 y. During the decay of 137Cs and its short-lived progeny 137mBa, both photons (0.662 MeV is most prominent) and beta particles (0.25 MeV) are emitted per nuclear transformation. Because most forms of 137 Cs are quite soluble in body fluids, it is readily absorbed from the lungs and GI tract after an inhalation or ingestion exposure. Cesium-137 reaching the systemic circulation behaves like a potassium analog and is distributed broadly throughout body organs and tissues where it is retained with a biological half-life on the order of 100 d. If the 137Cs becomes encapsulated in less soluble matrices, retention time in the lungs will increase and thereby slow the transfer to other body organs, changing the relative distribution of dose to the lung relative to other body organs and tissues. Further information on the biokinetics and dosimetry of internally-deposited 137Cs can be found in ICRP Publication 71 (ICRP, 1995a), Leggett et al. (2003), and NCRP Report No. 161 (NCRP, 2008a).

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Treatment

Decorporation treatment for internal cesium contamination is with the oral administration of Prussian blue insoluble (Radiogardase®). Prussian blue enhances excretion of isotopes of cesium and thallium from the body by means of ion exchange. Orally administered Prussian blue traps cesium in the gut, interrupts its reabsorption from the GI tract, and thereby increases fecal excretion. Treatment should begin as soon as possible after contamination is suspected. However, treatment is still effective if delayed (NCRP, 2008a). The dose of insoluble Prussian blue for adults and adolescents is 3 g orally three times daily. Doses up to 20 g d–1 for significantly contaminated adults have been administered. The dose of insoluble Prussian blue for pediatric patients 2 to 12 y of age is 1 g orally, three times daily for 30 d (FDA, 2010a). Dosing has not been well established for infants 0 to 2 y of age, even though these young ages would constitute the most radiosensitive group (Table 9.3). Radiogardase® has been FDA approved for the treatment of internal contamination with radioactive cesium and thallium except for infants <2 y of age (FDA, 2003). For patients who are unable to swallow capsules, the capsules may be opened and mixed with bland food or liquids. Experience has been with 137Cs ingestion, and the efficacy of Prussian blue in an inhalation incident is untested (NCRP, 2008a). 9.6.3

Medical Follow-Up After Treatment

Use whole-body counting to evaluate excretion kinetics and to assess need for further treatment (NCRP, 2008a). Gamma-ray counting of urine and fecal samples will also provide useful information. 9.7 Medical Management of a Cobalt-60 Intake 9.7.1

Overview

Cobalt-60 has a radioactive half-life of 5.3 y. During the decay of Co, high-energy photons (1.17 and 1.33 MeV) and low-energy beta particles (0.097 MeV per nuclear transformation) are emitted. The retention of inhaled 60Co in the lungs and its absorption from the GI tract will depend on the in vivo solubility of the form(s) inhaled or ingested. Exposures to 60Co may also occur from shrapnel wounds. Cobalt-60 reaching the systemic circulation after absorption from the lungs or GI tract is expected to deposit in the liver and, to a lesser extent, broadly in other body organs. Further information on the biokinetics and dosimetry of 60Co can be found in ICRP Publication 67 (ICRP, 1993) and NCRP Report No. 161 (NCRP, 2008a).

60

Mechanism of action: ion exchange.

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

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

Notes

Adults and adolescents: 3 g orally three times daily.

Administer

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.

Contraindications

TABLE 9.3—Prussian blue treatment (NCRP, 2008a).a

Experience with Prussian blue has been entirely in ingestion incidents. Likely not as effective in dose reduction for an incident involving a relativelyinsoluble form of 137Cs. Up to weekly assessment by whole-body counting or urine and fecal bioassays to assess treatment efficacy.

Monitor Clinically

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Pregnancy Category Cb

Continue for a minimum of 30 d per FDA guidance.

b

FDA (2010a). Pregnancy categories are defined in Appendix J.

a

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

Infants (not FDA approved): Consider benefit versus risk of a scaled dose.

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138 / 9. MANAGEMENT OF INTERNALLY-CONTAMINATED PERSONS 9.7.2

Treatment

There are no FDA approved treatments for absorbed cobalt. For high internal depositions from an inhalation exposure, off-label use of Ca- or Zn-DTPA (Table 9.2) and calcium ethylenediaminetetraacetic acid (Ca-EDTA) [edetate calcium disodium injection, calcium disodium Versenate®; do not confuse with edetate disodium, Endrate® (Cerner Multum, Inc., Denver, Colorado) (FDA, 2008)] (Table 9.4) may have potential use, although there is no clinical experience in use of either DTPA or EDTA for cobalt inhalation incidents. If DTPA is considered, the same regimen as described for 241 Am is recommended. The first dose can be very important. Clinicians considering using these treatments should consult with a knowledgeable physician experienced with treating internallydeposited radionuclides. It is also possible that the 60Co may be present as shrapnel in a wound of a person near an RDD detonation. See NCRP (2006) for additional guidance on contaminated wounds. 9.7.3

Medical Follow-Up After Treatment

Obtain whole-body counting to evaluate excretion kinetics and to assess need for further treatment (NCRP, 2008a). Gamma-ray counting of urine and fecal samples will also provide useful information. 9.8 Medical Management of an Iodine-131 Intake 9.8.1

Overview

Iodine-131 is likely to be the dominant initial internally-deposited radionuclide after a nuclear reactor or other incident involving fresh fission products. Iodine-131 has a radioactive half-life of 8 d. During the decay of 131I, photons (0.364 MeV and others) and lowenergy beta particles (0.192 MeV per nuclear transformation) are emitted. Most inhaled or ingested forms of radioiodines are rapidly absorbed into the systemic circulation from the lungs or the GI tract. Approximately one-third of the absorbed radioiodine is deposited in the thyroid and the remainder is excreted in urine, although this apportionment is highly variable. Further information on the biokinetics and dosimetry of internally-deposited 131I can be found in ICRP Publication 56 (ICRP, 1989), and NCRP Report No. 159 and No. 161 (NCRP, 2008a; 2008b). Extensive clinical experience with the health effects of 131I was garnered after the 1986 Chernobyl nuclear reactor accident. Children are more sensitive to the carcinogenic effects of radioactive iodine compared to adults (NCRP, 2008b).

Pregnancy Category Bb

Notes

b

RxList (2010). Pregnancy categories are defined in Appendix J.

a

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

Ca-EDTA 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 intramuscular divided into equal doses spaced 8 to 12 h apart. Administer for 1 d as above unless medically indicated for extended treatment.

Administer

Patients with active renal disease or hepatitis.

Contraindications

TABLE 9.4—Ca-EDTA treatment (NCRP, 2008a).a

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

Monitor Clinically 9.8 MEDICAL MANAGEMENT OF AN IODINE-131 INTAKE

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140 / 9. MANAGEMENT OF INTERNALLY-CONTAMINATED PERSONS 9.8.2

Treatment

Iodine-131 requires urgent intervention to be effective. Thyroid blockade by administration of KI can substantially reduce thyroid uptake of radioactive iodine if administered 1 to 2 d before exposure or within 4 to 5 h after exposure. Effectiveness diminishes rapidly with time after uptake of 131I (Nauman and Wolff, 1993). Administration 16 h or more after exposure may have little protective effect. KI is FDA-approved for the treatment of internal contamination with radioiodine, and FDA has developed treatment guidance based on age, predicted thyroid exposure, pregnancy, and lactation status (FDA, 2001) (Tables 9.5 and 9.6). Children younger than 18 y of age and pregnant women should receive priority for treatment due to increased sensitivity to radioactive iodine of children and fetuses. Recommendations for solution preparation using 130 mg tablets (Table 9.7) as well as the use of commercial KI preparations, such as ThyroShield® (Fleming and Company, Pharmaceuticals, Fenton, Missouri) 65 mg mL–1 solution, Iosat® (Anbex, Inc., Williamsburg, Virginia) 130 mg tablets, or ThyroSafe® (Recipharm, Stockholm) 65 mg tablets, that have been approved by FDA (2006b; 2006c; 2010b). Pharmacies may stock other forms of stable iodine useful in an emergency, such as a saturated solution of KI and Lugol’s solution. KI should be administered with caution to patients with a history of thyroid disease, iodine hypersensitivity, dermatitis herpetiformis, and hypocomplementemic vasculitis (FDA, 2001). 9.8.3

Medical Follow-Up After Treatment

Repeated thyroid counts will provide data needed to calculate the absorbed dose to the thyroid from 131I that was not blocked from deposition in the thyroid. 9.9 Medical Management of an Iridium-192 Intake 9.9.1

Overview

Iridium-192 has a radioactive half-life of 74 d. During the decay of 192Ir, photons (0.317 MeV and others) and low-energy beta particles are emitted. The retention of inhaled 192Ir in the lungs and its absorption from the respiratory tract will depend on the in vivo solubility of the form(s) inhaled or ingested. Iridium-192 reaching the systemic circulation after absorption from the lungs or GI tract is expected to deposit preferentially in the liver, kidneys and spleen with lower concentrations in other organs. Further information on the biokinetics and dosimetry of internally-deposited 192Ir can be

KI Dose (mg)c

130 130 130

65 65 32 16

≥5 (500) ≥0.1 (10) ≥0.05 (5) ≥0.05 (5) ≥0.05 (5) ≥0.05 (5) ≥0.05 (5)

Adults >40 y

Adults 18 – 40 y

Pregnant or lactating women

Adolescents 12 – 18 yc

Children 3 – 12 y

1 month – 3 y

Birth – 1 month

0.125

0.25

0.5

0.5

1

1

1

Number of 130 mg Tablets

0.25

0.5

1

1

2

2

2

Number of 65 mg Tablets

0.25

0.5

1

1

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. b Without KI treatment. c Adolescents approaching adult size (>70 kg) should receive the full adult dose (130 mg).

Age Category

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

TABLE 9.5—Threshold thyroid radiation doses and recommended doses of KI for different risk groups (adapted from FDA, 2001).a 9.9 MEDICAL MANAGEMENT OF AN IRIDIUM-192 INTAKE

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Additional therapy modes (off-label): saturated solution of KIa (1 g mL–1), 5 or 6 drops in juice.

KI orally, per FDA guidance (Table 9.5).

Administer

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

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

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

Contraindications

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

Notes

Neonates especially should be monitored for transient hypothyroidism by measurement of thyroid-simulating hormone.

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

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

Monitor Clinically

TABLE 9.6—KI treatment for intakes of radioactive iodine (adapted from FDA, 2001; 2010b; NCRP, 2008a).

142 / 9. MANAGEMENT OF INTERNALLY-CONTAMINATED PERSONS

Pregnancy Category D.c Agranulocytosis is a potential side effect.

Propylthiouracil,d 50 mg tablets.

b

WebMD (2010). RxList (2010). c Pregnancy categories are defined in Appendix J. d Drugs (2010).

a

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

Methimazole®b [Tapazole® (Cerner Multum, Inc., Denver, Colorado)], 5 and 10 mg tablets. Drug hypersensitivity and nursing women.

Drug hypersensitivity and nursing women.

Monitor complete blood count during treatment.

Monitor complete blood count during treatment. 9.9 MEDICAL MANAGEMENT OF AN IRIDIUM-192 INTAKE

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144 / 9. MANAGEMENT OF INTERNALLY-CONTAMINATED PERSONS TABLE 9.7—Preparation procedure for emergency administration of KI tablets to infants and children using 130 mg tablets (adapted from FDA, 2006b). INSTRUCTIONS 1. Grind the potassium iodide tablet into powder. Put one (1) 130 mg potassium iodide tablet into a small bowl and grind it into a fine powder using the back of the metal teaspoon against the inside of the bowl. The powder should not have any large pieces. 2. Dissolve the potassium iodide powder in water. Add four (4) teaspoonfuls of water to the potassium iodide powder in the small bowl. Use a spoon to mix them together until the potassium iodide powder is dissolved in the water. 3. Mix drink of choice with potassium iodide powder and water solution. Add four (4) teaspoonfuls of the desired drink to the potassium iodide powder and water mixture described in Step 2. HOW MUCH OF THE POTASSIUM IODIDE (KI) MIXTURE TO GIVE A CHILD USING 130 mg PREPARATIONS The number of teaspoonfuls of the drink to give a child depends on the child’s age. The chart below tells you how much to give a child. You should give potassium iodide once a day until a risk of significant exposure to radioiodines (radioactive iodine) no longer exists.

If Your Child is:

Give Your Child this Amount of KIa Drink

A teenager between 12 and 18 y of ageb

4 teaspoonfuls

Between 4 and 12 y of age

4 teaspoonfuls

Over 1 month through 3 y of age

2 teaspoonfuls

An infant from birth through 1 month of age

1 teaspoonful

HOW ALREADY PREPARED POTASSIUM IODIDE (KI) MIXTURE SHOULD BE STORED • Potassium iodide mixed with any of the recommended drinks will keep for up to 7 d in the refrigerator. • Potassium-iodide drink mixtures should be prepared fresh weekly; unused portions should be discarded. a

You should give your child one dose each day as directed. Teenagers approaching adult size [≥70 kg (≥154 pounds)] should receive the full adult dose (130 mg tablet or 8 teaspoonfuls of KI mixture). b

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found in ICRP Publications 30 and 68 (ICRP, 1980; 1994b) and NCRP Report No. 161 (NCRP, 2008a). 9.9.2

Treatment

There have been very few recorded intakes of 192Ir. Therefore, there is no documented information about the efficacy of various agents for removing internally-deposited 192Ir. For a case in which there has been a high intake of 192Ir, off-label use of DTPA (Table 9.2) may be considered (NCRP, 2008a). Clinicians considering using this treatment should consult with a knowledgeable physician experienced with treating internally-deposited radionuclides. It is also possible that the 192Ir may be present as shrapnel in a wound of a person near an RDD detonation. Metal fragments, if present, could potentially deliver high doses and removal of fragments must be paramount in the treatment of individuals with fragments in clothing or embedded in wounds. See NCRP Report No. 156 (NCRP, 2006) for additional guidance on contaminated wounds. 9.9.3

Medical Follow-Up After Treatment

Direct whole-body counting methods are preferred. Photon counting of weekly or monthly fecal analysis may be useful to document excretion. Urine analysis may have limited application due to little absorption to blood. 9.10 Medical Management of a Plutonium-238 or -239 Intake 9.10.1 Overview Like americium and uranium, plutonium is a member of the actinide series, those heavy elements beginning with actinium and ending with the heaviest known elements. The radionuclides of primary interest here, 238Pu and 239Pu, have radioactive half-lives of ~88 and 24,000 y, respectively. During the decay processes, these radionuclides emit alpha particles (5.49 and 5.15 MeV, respectively) and low-energy x rays. The retention of inhaled 238Pu and 239 Pu in the lungs and their absorption from the GI tract will depend on the in vivo solubility of the form(s) inhaled or ingested. Plutonium reaching the systemic circulation after absorption from the lungs or GI tract is deposited primarily in the liver and skeleton where it is avidly retained. Further information on the biokinetics and dosimetry of internally-deposited plutonium radionuclides can be found in ICRP Publications 67 and 68 (ICRP, 1993; 1994b) and NCRP Report No. 161 (NCRP, 2008a).

146 / 9. MANAGEMENT OF INTERNALLY-CONTAMINATED PERSONS 9.10.2 Treatment Treatment may be divided into three categories based on route of exposure. Plutonium contamination of intact skin is easily managed by washing with water or detergents (NCRP, 2008a). Areas not completely cleaned may be covered after loose material is removed, and the individual may be sent home. Contaminated wounds or burns require a more aggressive approach. Begin with simple decontamination and irrigation of the wound or burn. More extensive treatment by excision of wounds requires consultation with a physician familiar with treating internal depositions of plutonium. Treatment of these wounds may require initial chelation therapy with DTPA (Table 9.2) prior to surgical excision of the wound to limit systemic absorption. In burns, much of the plutonium contamination may slough off when the eschar does (NCRP, 2006). Ca- and Zn-DTPA are also the chelation therapy treatments of choice for inhalation incidents involving actinides including plutonium, americium and curium (NCRP, 2008a). Because the chemical form usually will not be known in the immediate aftermath of an incident, DTPA should be given as soon as possible after an inhalation exposure to chelate any plutonium in a relatively soluble form that is circulating in or available to, tissue fluids and plasma. Efficacy of DTPA is greatest if administered with 24 h of intake but can still be effective if given later (Marcus, 2004; NCRP, 2008a). An additional therapeutic approach to consider for high inhalation intakes of plutonium in a relatively-insoluble form is bronchoalveolar or lung lavage. This procedure, which requires general anesthesia, is discussed in greater detail in NCRP (2008a). 9.10.3 Medical Follow-Up After Treatment Ongoing urine bioassays should be used to evaluate the efficacy of treatment. 9.11 Medical Management of a Radium-226 Intake 9.11.1 Overview Radium, like calcium, strontium and barium, is a member of the alkaline earth elements. The radionuclide of interest here is 226Ra, which has a radioactive half-life of 1,600 y and emits alpha particles with a total energy of 4.8 MeV per nuclear transformation and gamma rays. A number of the radioactive decay products of 226Ra emit either alpha or beta particles and photons of different energies. The retention of inhaled 226Ra in the lungs and its absorption

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from the GI tract will depend on the in vivo solubility of the form(s) inhaled or ingested. Radium-226 reaching the systemic circulation after absorption from the lungs or GI tract will deposit preferentially in the skeleton where it is avidly retained. Further information on the biokinetics and dosimetry of internally-deposited 226Ra can be found in ICRP Publication 67 (ICRP, 1993) and NCRP Report No. 161 (NCRP, 2008a). 9.11.2 Treatment One or more of the drugs in Table 9.8 should be taken as soon as possible post-incident to block intestinal absorption (NCRP, 2008a). 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® (GlaxoSmithKline, London) ~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). 9.11.3 Medical Follow-Up After Treatment Ongoing urine bioassays should be used to evaluate the efficacy of treatment. 9.12 Medical Management of a Strontium-90 Intake 9.12.1 Overview Strontium, like calcium, barium and radium, is a member of the alkaline earth elements. The radionuclide of primary interest here is 90Sr, which has a radioactive half-life of 28.8 y. During the decay of 90Sr and its short-lived progeny 90Y, beta particles are emitted with a total energy of 1.1 MeV per nuclear transformation. The retention of inhaled 90Sr in the lungs and its absorption from the GI tract will depend upon the in vivo solubility of the form(s) inhaled or ingested. Strontium-90 reaching the systemic circulation after absorption from the lungs or GI tract will deposit preferentially in the skeleton where it is avidly retained. Further information on the biokinetics and dosimetry of internally-deposited 90Sr can be found in ICRP Publication 67 (ICRP, 1993) and NCRP Report No. 161 (NCRP, 2008a). 9.12.2 Treatment For most forms of strontium, bone surface is the dose-limiting organ for both inhalation and ingestion. Strontium titanate is an exception because its very insoluble nature leads to prolonged

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

Aluminum hydroxidea[Alternagel® (Merck and Company, Whitehouse Station, New Jersey), Amphojel® (American Home Products, Madison, New Jersey)] oral

Adults: 60 – 100 mL (1,200 mg) Pediatric: 50 mg kg–1, not to exceed adult dose (NCRP, 1980) or Barium sulfateb, 100 – 300 g oral in a single dose in 240 mL water Consider laxatives during first 24 h after strontium ingestion.

Dose to reduce intestinal absorption:

Notes

Administer

Hypersensitivity

Contraindications

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.

Monitor Clinically

TABLE 9.8—Drugs to reduce GI absorption of strontium or radium (NCRP, 2008a).

148 / 9. MANAGEMENT OF INTERNALLY-CONTAMINATED PERSONS

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

Calcium phosphate,d 1 to 2 tablets once

b

Drugs (2010), FDA (2010a). Drugs (2010), Mayo (2010), NLM/NIH (2010). c Nutrition Surplus (2010). d Drugs (2010).

a

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

Sodium alginatec, 5 g oral twice daily, then 1 g four times daily with water

Alternative regimens: 9.12 MEDICAL MANAGEMENT OF A STRONTIUM-90 INTAKE

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150 / 9. MANAGEMENT OF INTERNALLY-CONTAMINATED PERSONS retention of 90Sr in the lung. 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 most useful in the medical management of inhalation and ingestion cases with soluble forms of strontium (Tables 9.8 and 9.9): Drugs to reduce the absorption of strontium Drugs given to reduce the absorption of soluble forms of strontium that are given here and listed in Table 9.8 should only be given during the first 24 h after the radionuclide intake. Other treatments for absorbed strontium are presented later. • aluminum hydroxide: a dose of 60 to 100 mL to reduce intestinal absorption. • barium sulfate: 100 to 300 g orally in a single dose in 250 mL water. • alternative regimens: - sodium alginate, 5g oral twice daily, then 1 g four times daily with water. - calcium phosphate, 1 to 2 tablets, once. Drugs to reduce skeletal deposition of strontium The drugs listed in Table 9.9 should be started as soon as possible after the radionuclide intake, at least within 12 h of intake, if possible. The drugs listed in Tables 9.8 and 9.9 can be given together during the early post-intake period. 9.12.3 Medical Follow-Up After Treatment Ongoing urine bioassays should be used to evaluate the efficacy of treatment. 9.13 Medical Management of a Uranium-235 or -238 Intake 9.13.1 Overview Like americium and plutonium, uranium is a member of the actinide series, those heavy elements beginning with actinium and ending with the heaviest known elements. The radionuclides of primary interest here, 235U and 238U, have very long radioactive halflives, 7.0 × 108 y and 4.5 × 109 y, respectively. During radioactive decay, they both emit alpha particles with total energies of 4.4 MeV (235U) and 4.2 MeV (238U) per nuclear transformation and gamma rays. Both decays are followed by a number of further decays resulting in the emissions of additional alpha and beta particles. The

Strontium gluconate

Strontium lactate

IV

Calcium gluconatec ampules

600 mg daily for up to 6 d

IV

500 to 1,500 mg d–1

Oral

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

As directed

Oral

Route of Administration and Dose

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

Compound

Monitor serum calcium.

Too rapid an IV infusion of calcium gluconate may precipitate hypotension. Monitor blood pressure during administration.

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).

May cause constipation.

Remarks

TABLE 9.9—Drugs to reduce skeletal deposition of strontiuma (do not use for chromium decorporation) (NCRP, 2008a).

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May be beneficial, limited clinical experience, potentially toxic (NCRP, 1980)

IV

Route of Administration and Dose

b

Also see case studies in Section 20.21 of NCRP Report No. 161 (NCRP, 2008a). Drugs (2010). c RxList (2010).

a

Consider parathormone

Compound

TABLE 9.9—(continued) Remarks

152 / 9. MANAGEMENT OF INTERNALLY-CONTAMINATED PERSONS

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retention of inhaled 235U and 238U in the lungs and their absorption from the GI tract will depend upon the in vivo solubility of the form(s) inhaled or ingested. Uranium reaching the systemic circulation after absorption from the lungs or GI tract will deposit preferentially in the skeleton like an alkaline-earth element. Because of the low specific activity of 238U, natural uranium, or DU, relatively large masses of uranium are associated with relatively low levels of activity. Therefore, heavy-metal toxicity to the kidneys may be more important for large intakes of these radionuclides than the associated radiation exposure (Section 7.2.5). Further information on the biokinetics and dosimetry of internally-deposited uranium radionuclides can be found in ICRP Publication 69 (ICRP, 1995b), Leggett (1994), and NCRP Report No. 161 (NCRP, 2008a). 9.13.2 Treatment For an intake of uranium, treatment is primarily by alkalinization of urine. Alkalinization of the urine promotes excretion of uranium-bicarbonate, and reduces the probability of acute tubular necrosis. Use isotonic sodium bicarbonate, 250 mL slow IV infusion or two bicarbonate tablets (650 mg per tablet) orally every 4 h until the urine reaches a pH of eight to nine. Continue treatment for 3 d. Blood pH and serum electrolytes should be monitored. A pre-existing hypokalemia may be unmasked and this treatment should be used with caution in congestive heart failure or in disease states with sodium retention. For a high level intake of uranium, consider off-label dosing as follows: Etidronate [Didronel® (Warner Chilcott, Dublin)] 400 mg orally once a day for 3 d or Diamox® (Teva Pharmaceuticals, North Wales, Pennsylvania) 500 mg orally twice a day for 3 d (Table 9.10). 9.13.3 Medical Follow-Up After Treatment Ongoing urine bioassays should be used to evaluate the efficacy of treatment.

Pregnancy Category Cb

For high level of uranium intake, consider off-label diuretic dosing as follows: Etidronate (Didronel®a – 400 mg orally daily, or Diamox®a – 500 mg orally two times per day. Continue × 3 d. Pregnancy Category C

b

Drugs (2010). Pregnancy categories are defined in Appendix J.

a

Consider dialysis for suspected high levels of intake.

Alkalinization of the urine promotes excretion of uranium-carbonates, reduces probability of ATN. Alkalosis has occurred after one tablespoonful in a young infant.

Isotonic sodium bicarbonatea, 250 mL slow IV infusion, or two bicarbonate tablets oral q 4 h until the urine reaches a pH of 8 – 9. Continue × 3 d.

Pregnancy Category C

Notes

Administer

Anuria

Contraindications

TABLE 9.10—Uranium treatment group.

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

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

Monitor Clinically

154 / 9. MANAGEMENT OF INTERNALLY-CONTAMINATED PERSONS

10. Social, Psychological and Communication Issues Associated with Screening and Monitoring a Population 10.1 Introduction As noted in other parts of this Report, there will be numerous technical and logistical issues associated with a large-scale population screening and monitoring effort. Yet, in the end, understanding and addressing the human dimension may pose some of the biggest challenges. Incidents involving radiation and other toxic hazards have the capacity to upset people in new and unusual ways. Indeed, people find such agents “a good deal more threatening than both natural hazards of even the most dangerous kind and mechanical mishaps of considerable power” (Erikson, 1994). A population screening and monitoring program launched in the aftermath of a radiological or nuclear incident will likely find many people in the affected areas anxious and distressed. Concerns and uncertainties about potential health impacts will be strong and pervasive. Furthermore, because technological disasters are seen as preventable, there could be a high level of anger and mistrust, and a powerful sense of violation and betrayal (Becker, 1997; 2001a). A monitoring program undertaken after an RDD or IND will likely find the population even more severely affected than they would be following a natural disaster. The shock of the attack will still be strong and dreadful images such as those seen after the Oklahoma City bombing and the September 11, 2001 attacks may be fresh in people’s minds. Some affected individuals will have suffered the death of loved ones. Others may still be waiting for word about missing relatives. Families will be deeply worried about their children and women who are expecting babies will be anxious about their pregnancies. Depending on the situation, there may also be concerns about environmental contamination and the safety of food and water supplies. In addition, there may be a continuing threat and fear of possible additional attacks. In summary, people will be shaken and on edge. 155

156 / 10. SOCIAL, PSYCHOLOGICAL AND COMMUNICATION ISSUES 10.2 General Approach If a population screening effort is undertaken after a large-scale incident or a terrorist attack, it should be sensitive in the way that it is organized and conducted. Care should be taken to ensure that an activity fundamentally aimed at helping people does not alienate or further traumatize or stigmatize them. Far from being just a technical, logistical or medical activity, a population screening program will be one of the main points of contact between government agencies and the population affected by a calamity. It should address people’s concerns, be a source of clear information, foster resilience, build trust and confidence, and demonstrate in action the concern that the nation has about its citizens. To do so, every aspect of the planning and implementation of a population screening program should be informed by an awareness of social, psychological and communication issues (Becker, 2001a; 2001b). 10.3 Program and Center Names The term population monitoring itself may need to be reconsidered in some contexts. Although the term is familiar to health physicists, radiation agencies and some other professional audiences, it is not likely to be clear to many other professionals or members of the public (Becker, 2004). In fact, the word monitoring could even sound ominous to some. Perhaps more importantly, the term “population monitoring” does not effectively communicate what the real aim of these efforts is: helping to protect people’s health. In light of this, it may be worth considering alternate terms for population monitoring programs and centers. Certainly, an alternative such as community reception center (CDC, 2007) is far more effective than population monitoring center in accurately and clearly communicating the health protection emphasis of the effort. Ideally, the precise choice of terms should be informed by prior research and testing, community input, and an understanding of the cultural context. 10.4 Enlisting the Public as a Partner For population screening efforts to be effective, they should involve stakeholders as early as possible. First, community involvement is vital for establishing trust and showing clearly that local concerns are a priority. Particularly in a situation (e.g., after an incident) where the background level of anger and mistrust may already be high, the building of trust and confidence will be essential. Second, even the most comprehensive templates and protocols should be complemented by an understanding of local conditions, customs and cultural sensitivities. Third, only by tapping local

10.6 PROACTIVE APPROACH

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insights can outreach to the affected population be fully effective. For example, stakeholders may be able to identify ways of providing information to harder-to-reach portions of the population (e.g., the homeless). Planning for population screening and monitoring should incorporate provisions for establishing stakeholder advisory boards or other mechanisms for making members of the public a partner. 10.5 Communicating with Members of the Public Effective communication with the public will be crucial to every aspect of a population monitoring effort, from explaining what it is and how it works to successfully reaching the appropriate population. Because population monitoring is fundamentally about health and the protection of people’s health, agencies and spokespersons with a high level of credibility on health issues need to be at the heart of public communication efforts. Research on terrorism situations involving radiation, hazardous chemicals, or biological agents has shown that people’s concerns and questions tend to revolve around health issues (Becker, 2004; Glik et al., 2004; Henderson et al., 2004; Wray and Jupka, 2004). Furthermore, public opinion research has shown that when people are asked who they would trust to provide “accurate and reliable information about what is happening and what to do in the event of a terrorist attack,” they rank health-related professionals and agencies such as the CDC the highest (NCDP, 2007). For population monitoring programs, putting agencies and spokespersons with high credibility on health issues at the center of communication efforts will help to build trust and address the concerns that are foremost in people’s minds. 10.6 Proactive Approach The pressure of incidents is likely to leave little or no time to develop audiovisual materials, public service announcements, fact sheets, media packets, and other informational materials. Wherever possible, therefore, these should be developed in advance. In this regard, a “preincident message development” strategy may be useful. Increasingly utilized by CDC in the aftermath of the 2001 anthrax letter incidents, this involves performing research on the concerns, information needs, and preferred information sources of key audiences; utilizing the findings to prepare messages and other materials; and carefully testing them long before an actual incident

158 / 10. SOCIAL, PSYCHOLOGICAL AND COMMUNICATION ISSUES occurs (Becker, 2004; Vanderford, 2004). The virtue of this approach is twofold: first, it means that messages and informational materials are already available when an emergency occurs, enabling agencies to move rapidly; and second, it means that these materials have already been tested and revised (e.g., identifying and removing unclear or ambiguous terms), increasing the likelihood of successfully transmitting critical information to the public or other audiences. Among the materials that could be developed and tested in advance for a population monitoring effort are audiovisual materials explaining the program to members of the public; question and answer web content aimed at addressing the most common questions and concerns; audiovisual materials on “what to expect” that could be shown to people waiting in line at a reception center; and follow-up fact sheets, DVDs or other items that could be taken home after people leave the center. Materials should also be developed in advance to address some of the questions likely to arise in relation to pets. For many individuals, pets are thought of as part of the family, and concerns about their well-being can significantly affect people’s decisions and actions during a disaster (Hall et al., 2004). Informational materials should be available to address questions such as whether pets should be brought to reception centers and what steps can be taken to protect them. At the present time, no scientific consensus exists regarding the screening and treatment of pets for possible internal contamination. However, there is agreement that external decontamination is helpful and appropriate. Pet owners should be encouraged to wash their own pets when possible. Washing an animal with soap and water should remove a large majority of any external contaminants, thereby reducing any risk of harm to the pet and also reducing the likelihood of spreading contamination to the owner (CDC, 2007). Whenever possible, materials intended for use in population monitoring efforts should be developed and tested in multiple languages so that all parts of the population in an affected area can be reached. Although some informational materials prepared before an incident will require the addition of details specific to the incident, preparing the materials before an incident will provide a population screening and monitoring program with a strong, empirically grounded foundation for its communication efforts. 10.7 Information Hotlines Past experience with various emergencies suggests that there could be a massive demand for information from telephone hotlines (Vanderford, 2004). While general hotlines will have undoubtedly

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already been setup in connection with the incident, the population screening and monitoring program must have the capacity to field the large number of calls dealing specifically with screening or monitoring issues. Thus, agencies should have well-rehearsed plans, dedicated phone facilities and phone lines, and trained personnel to rapidly setup and operate such hotlines. Hotline arrangements (including mechanisms to ensure that accurate and up-todate information is available) should be regularly and realistically tested through exercises. 10.8 Special Services, Approaches and Materials for Children In planning and conducting a population monitoring program, the needs and well-being of children will require particular attention (Gurwitch et al., 2004; Markenson and Redlener, 2004). In any disaster situation, children have unique vulnerabilities. They may be exposed to the same frightening sights, sounds and smells that adults experience, but not have the coping skills, resources, emotional maturity, and life experience to understand and deal with what is going on around them. A terrorist attack involving radioactive material could be particularly distressing, since it could involve mass casualties, invisible contamination, horrifying and grotesque images, and perhaps evacuation and dislocation. For some children, it could also involve being injured or made ill, being separated from parents, or seeing parents, siblings, teachers or schoolmates injured, sickened or killed. Depending on the nature and location of a terrorist attack, a substantial proportion of the people being seen at a reception center could consist of children. Indeed, with recent historical experience showing that some terror attacks specifically target schools or other venues frequented by children (Brandenburg and Regens, 2006); it is even possible to envision scenarios where a majority of those seen in a population monitoring effort are children. It will be important, therefore, for population monitoring efforts to be as child-friendly as possible. The last thing a population monitoring program should be doing is further traumatizing children who are already likely to be deeply distressed by a terrorist incident. One important consideration should be to find a location where children will be as comfortable as possible. Some communities may be able to make use of existing pediatric healthcare facilities, such as a children’s hospital. In other situations, communities may opt to utilize a school or daycare facility. “Children spend the majority of their waking hours at school or in a child care setting. These settings are familiar and comfortable to children, and generally are

160 / 10. SOCIAL, PSYCHOLOGICAL AND COMMUNICATION ISSUES experienced as safe, secure environments. As such, school and child care settings are excellent locations for working with children before, during, and after a disaster” (Gurwitch et al., 2004). On the other hand, some communities may choose to avoid the use of schools or childcare centers because of concerns about stigma or other issues. Ultimately, such decisions should rest with people in the affected area. Regardless of the specific location chosen, reception centers should have dedicated areas set aside for families with young children. To reduce levels of fear and distress, children and parents or guardians should be kept together. Wherever possible, the dedicated areas for families should be decorated in such a way as to make young children feel comfortable. For example, bright colors, children’s toys and furniture, familiar cartoon characters, etc. could be used. Program staff members who will be interacting with children should be given specific training on children’s mental-health issues. Similar to what is often done in pediatric wards and children’s hospitals, staff could also wear clothing with children’s characters or other icons that help children feel comfortable. In addition, because some of the equipment that is used in population screening or monitoring could be frightening to children, consideration should be given to developing partial covers with children’s characters on them (provided this does not interfere with the operation of the equipment). Finally, it will be useful to have age-appropriate informational materials, explanations, coloring books, videos, and other messages to help children and families understand and cope with the situation. Few such materials currently exist for situations involving radioactive materials, so it will be important for them to be developed as part of preincident planning efforts (Becker, 1997; 2002; Gurwitch et al., 2004; NCRP, 2001). 10.9 Persons with Reproductive and Fertility Concerns Special attention will also be needed for pregnant women and persons with reproductive and fertility concerns. Women who are expecting babies are likely to be extremely concerned about their pregnancies. Some may be contemplating abortion out of fear that exposure to radiation may have harmed the developing fetus. In addition, both men and women may have more general concerns about whether potential exposure to radioactive materials could affect fertility and future pregnancies. It will be important, therefore, for reception centers to have up-to-date information and counseling services available to discuss reproductive decisions and address questions and concerns (CDC, 2005b; ICRP, 2000).

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10.10 Other Special Populations Planning and implementation of a population monitoring program should be informed by an awareness of the needs and vulnerabilities of other special populations. For example, reception centers and their facilities should be made accessible to individuals with physical impairments, and appropriate informational messages and materials need to be available for persons who are blind or hard-of-hearing. Facilities in the community such as sports arenas and high school gymnasiums may be designed to meet the needs of special populations. In addition, consideration should be given to how the monitoring program can best accommodate persons with mental illness. More generally, it will be important for reception centers to have clinical staff onsite. With large numbers of people coming in for screening, it is likely that some individuals will arrive with illnesses, injuries, or other medical problems. Nurses, emergency medical technicians, doctors, or physicians assistants should be on hand to address any such medical problems that may arise. In addition, medical personnel, working in conjunction with mentalhealth professionals, should be available to discuss and explain test results, provide counseling when useful, and answer people’s healthrelated questions. 10.11 Addressing Staff Concerns and Information Needs Relatively few professionals are specialists in screening or monitoring a population. Thus, in the event of a large monitoring program having to be setup, it is likely that many other professionals from emergency-responder organizations, medicine, nursing, public health, health physics, and other fields will be needed to staff and operate centers. As recent research confirms, even experienced emergency responders and receivers may have significant concerns, questions and apprehensions about dealing with radiation and radioactivity (Becker, 2004; 2005). It will be important to ensure that all members of staff are provided with information that specifically addresses any such concerns. Materials prepared in advance of an incident may be especially useful, particularly if they are grounded in empirical research on the radiation-related concerns most commonly expressed by professionals. One useful component in the overall communication effort is to have a series of “just-in-time” videos available. Incorporating the most essential, practical information needed by staff, these brief videos could be viewed shortly before the commencement of a population monitoring program (or just-in-time). This would serve

162 / 10. SOCIAL, PSYCHOLOGICAL AND COMMUNICATION ISSUES both as a rapid, practical refresher and as a mechanism for addressing staff concerns and information needs. It is also crucial to remember that population monitoring efforts will depend not just on those who handle various monitoring procedures, but also on a host of other types of staff. These may range from support personnel to cleaning staff to maintenance workers. Many will have had little or no opportunity to become acquainted with radiation issues in their careers. Thus, it will be useful to have tailored videos and informational materials to answer their questions and address their health and safety concerns. 10.12 Staff Support In addition to providing information, it will be important to have adequate mental-health support available for all reception center staff. Working in a reception center could be highly stressful. The numbers of people coming to the center could be large, and many could be apprehensive and upset. Furthermore, for some members of the public, the reception center may be the only opportunity to directly interact with someone “official” about the incident. Thus, staff may find themselves the focus of a host of questions and intense emotions. Indeed, in the aftermath of an incident, members of staff could encounter distrust or even outright hostility. This is particularly the case if “government” is seen as somehow involved in, responsible for, or negligent in relation to the disaster (Becker, 1997). Regardless of the type of incident, the combination of potentially long work hours, fatigue, extended periods of time away from home, and possible concerns about their own families could put reception center personnel at increased risk for emotional distress. Thus, all population monitoring programs should include a robust mental-health component for staff. This should include the provision of self-care and stress-management information, regular rest breaks, buddy/peer support arrangements, and support groups. 10.13 Training Exercises In designing and conducting training exercises related to population monitoring, it will be important to fully incorporate social, psychological and communication issues (Becker, 2001b). All of the areas identified above are crucial to the success of a population monitoring program; as such, they need to be explicitly addressed and practiced on a regular basis. Without adequate consideration of relevant social, psychological and communication issues, exercises will be unrealistic and of limited value in preparing agencies and responders to deal with the complex challenges associated with population monitoring after a large-scale accident or terrorist attack.

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10.14 Understanding and Addressing Responder Concerns and Information Needs Emergency responders will be at the front line in any effort to manage a radiological or nuclear incident. To ensure their safety and enable them to perform their vital work, it is essential that responder concerns and information needs be understood and addressed. Furthermore, because emergency responders are often looked to by the public for information, it is important that responder agencies and emergency response professionals have the information they need to effectively speak to public concerns (Becker, 2010). Studies from several disciplines, which have employed various research methods, have begun to provide important insights into responder views, concerns, and information needs related to radiological or nuclear terrorism. One clear and consistent finding is that responders of all types have a high level of dedication to duty and a strong commitment to professional responsibilities. In focus groups and interviews, responders emphasize that their work is not just about doing a job; it is also about a powerful devotion to duty, helping and service (Becker, 2010). At the same time, it is also evident from focus groups, interviews and surveys that many responders have deep concerns regarding radiological or nuclear terrorism. These situations are seen as somehow “different” and many responders have serious doubts about individual and organizational readiness for responding to this “new” challenge. Although first responders appear to have a higher level of confidence than either public-health workers or hospital-based healthcare providers, all responder groups express very significant preparedness concerns. In addition, responders express a lower “comfort level” with radiation as compared to many other threats, and for some this perceived newness and lack of familiarity translate into greater apprehension than for other hazards (Becker, 2009; 2010; Becker and Middleton, 2008). Indeed, a strong and consistent finding from a series of survey research studies is that responders express a lower willingness to be involved in dealing with radiological or nuclear incidents than with most (or sometimes even all) other types of incidents. This is apparent in studies of the views of emergency medical service providers (DiMaggio et al., 2005), nurses (O’Sullivan et al., 2008; Veenema et al., 2008), physicians and nurses (Lanzilotti et al., 2002), healthcare workers (Qureshi et al., 2005), and hospital employees (Cone and Cummings, 2006). For example, Cone and Cummings (2006) examined the views of 1,711 hospital personnel in five states. A large majority, 87 %, indicated a willingness to work in response to a fire/rescue/

164 / 10. SOCIAL, PSYCHOLOGICAL AND COMMUNICATION ISSUES collapse mass casualty incident. For natural disasters, the figures were similarly high: snowstorm, 83 %; flood, 81 %; earthquake, 79 %; hurricane, 78 %; tornado, 77 %; ice storm, 75 %. For a flu epidemic, 72 % indicated a willingness to respond. When asked about man-made emergencies, however, the expressed willingness to work dropped significantly: biological or chemical incident, 58 %; radiation incident, 57 %. It is not clear exactly how this might play out in a real-world situation, since what people express in focus groups and surveys may not always mirror precisely what occurs in an actual incident. In a real-world emergency, the powerful commitment to duty that motivates responders could very well translate into higher levels of willingness to report than have been expressed in research studies. At a minimum, however, the research findings indicate that responders have deep concerns about situations involving radiation, and that these concerns have the potential to dramatically increase stress on responders and to make it harder for them to do their jobs. The findings also suggest at least the possibility that responder concerns (if unaddressed) could result in a reduced capacity for agencies to respond to a large-scale radiological or nuclear incident (Becker, 2010). Thus, it is essential that agencies work to better understand and address the concerns and information needs of first responders, health department personnel, hospital-based clinicians, and other healthcare staff. Initiatives to seek direct input from the front lines will be an essential part of this process as will additional research aimed at improving the effectiveness and responsiveness of messages and materials for responders.

11. Long-Term Follow-Up of Individuals 11.1 Identification of the Population to be Followed Radiological incidents may range from those in which few people have been contaminated externally to those in which large numbers of people are contaminated externally and internally. Clearly, the scope and characteristics of the official response will be dictated by a number of factors, including the type of incident and the nature of the release, the number of people potentially exposed, the geographic area and setting, and the overall availability of resources (e.g., personnel, facilities and expertise). It is not the intent of this Report to provide a detailed set of uniform recommendations and specific procedures that should be followed for every incident. Each will require a response tailored to the specific circumstances. Instead, this Report is intended to provide a more general set of guidelines that will help direct authorities to plan for, and respond to, a radiological or nuclear incident in their jurisdiction. Immediately following a radiological incident, it will be important first to identify and screen persons exposed or potentially exposed to radioactive materials from the incident, in order to assess external and internal contamination as discussed in Section 8. Rapid screening of individuals should include the administration of a brief and simple questionnaire that collects basic demographic information, the location of the person at the time of the incident and their subsequent movements, and a description of any precautions taken to reduce or avoid exposure. The primary objective of such a screening program is to identify individuals who may have been sufficiently exposed to warrant medical intervention. In many, if not most, places this may be all the local authorities can realistically accomplish. A second important step, if resources are available, would be to identify individuals who may have been exposed to radioactive materials or may have been contaminated at levels below those that require immediate medical attention, but who might benefit from longer-term follow-up to look for early signs or symptoms of radiation-induced disease. Identification of these individuals is a considerably more complex task, and may be beyond the capabilities of most hospitals or healthcare facilities. 165

166 / 11. LONG-TERM FOLLOW-UP OF INDIVIDUALS A third step that probably is well beyond the capabilities of most communities would be to establish a population (or cohort) of persons with a wide range of exposure and demographic characteristics that are registered in a central database and that can be monitored and followed for a longer period of time to ascertain adverse health effects. Such a comprehensive screening effort could provide the basis for a registry of persons potentially exposed to radiation from the incident. In most incidents, operation of this registry would require state or federal resources. Once intakes of radionuclides have occurred, it will be necessary to identify them and to determine whether quantities in the body exceed the CDG (Section 7). The next step will be to identify members of the exposed population who have sufficient quantities of radionuclides in their bodies to warrant follow-up using bioassays. NCRP recommends a bioassay action-level or benchmark equal to one-half of the Clinical Decision Guide (CDG) value for identifying those individuals for whom continued bioassays should be considered for the purpose of correlating internal contamination with long-term biological effects. The CDG, as defined in Section 7, provides a decision level that can be used uniformly as a basis for possible public screening and follow-up guides. Both direct (in vivo) and indirect (in vitro) bioassay methods should be used to determine not only which individuals require additional measurements, but also to identify those for whom the possible application of medical procedures (such as the administration of chelating agents to reduce the amounts of internally-deposited radionuclides) should be considered. Section 8 describes in more detail the bioassay procedures used to determine the quantities of radionuclides that have been absorbed and deposited within the body, and Section 9 discusses medical treatments. It is important to keep in mind that the CDG is a decision-aiding tool to guide authorities in actions taken following a single or a series of radiological or nuclear incidents. It should be applied in a flexible manner, especially in terms of the available treatment center resources and the number of individuals requiring care. Other factors that should be considered as well include the age of each individual who was exposed and the effective half-life of the radionuclide. For example, an internal radionuclide deposition corresponding to 1 CDG might not lead to decorporation treatment for people older than 50 y of age, particularly if the radionuclides have a long effective half-life.

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In the event that intakes involve multiple radionuclides, the following normalization method can be used to obtain a new CDG equivalent value: k

CDG mix =

Ai

∑ ----------------CDG

i=l

(11.1)

i

where Ai and CDGmix represent the intake for radionuclide i and the total number of CDG values in a mixed intake of all radionuclides, k. In essence, i is a numerical index for integer values ranging from 1 to k (Section 7.2.4). Any accompanying injuries must also be taken into account when deciding whether to conduct long term follow-up, particularly when the individual has significant exposures to other toxic chemical agents, or has multiple traumatic injuries (such as burns). Under these circumstances, the 0.5 CDG threshold for bioassay might need to be adjusted (i.e., reduced or increased), taking into consideration the multiple injuries or life threatening conditions. A history of relatively-high occupational exposures to long-lived radionuclides that could be retained in an individual’s body should also be considered. Although such situations may be rare, where exposure can be documented, the 0.5 CDG threshold may need to be reduced by an appropriate amount to account for the added contribution to dose from occupational exposures. Procedures for treatment of individuals with internally-deposited radionuclides are intended to reduce absorbed radiation dose and the risk of future adverse health effects. These procedures are focused on reducing absorption and internal deposition, or enhancing elimination of the absorbed radionuclide. The physician’s decision regarding possible treatment of an individual should be guided, in part, by knowledge of the patient’s exposure as outlined more fully in Section 8. Cessation of treatment should depend largely on the effectiveness of the specific treatment and be balanced with anticipated long-term risks associated with the exposure and risks or other disadvantages of continuing treatment. In the initial screening process, individuals should be classified according to levels of estimated internal contamination. The type and magnitude of the incident will affect the number of individuals that may need to be monitored and followed, given the resources available. For example, a device that widely disperses radioactive material may result in a very different need to monitor and follow exposed individuals than a point source that is relatively localized.

168 / 11. LONG-TERM FOLLOW-UP OF INDIVIDUALS 11.2 Classification of Persons to be Monitored or Followed The initial screening of individuals potentially exposed during the incident will enable responders to prioritize further action for each individual. Persons with the highest levels of internal radionuclide deposition would be referred to specialists for prompt, accurate assessment with more sophisticated equipment, and, if appropriate, medical treatment as described in Sections 8 and 9. These individuals would receive the highest priority in terms of immediate additional measurements and follow-up. This group would include individuals with high likelihood of intake, those who were at the site of the incident or nearby, and those exhibiting symptoms of early effects from radiation exposure. After medical treatment of patients with intakes >1 CDG, patients would receive bioassay measurements for a period of time that would depend on the type and amount of internal contamination. Individuals with intermediate amounts of contamination, both external and internal, would receive similar priority. This group would include those with moderate intakes outside of a geographic perimeter that defines the highest contamination levels. Such individuals would receive bioassay measurements if their estimated intakes exceeded 0.5 CDG. The duration of their participation in a monitoring program would depend on the amount of internal contamination and the resources available. Due to the large uncertainty in bioassay measurements, the decision to terminate bioassays should be made on a case-by-case basis. Nonetheless, dose rate limits for terminating follow-up activities for individual victims of an incident might be based on a sliding scale of between 1 and 5 mSv y–1 total effective dose. Using this approach, the extent to which an individual would be actively followed and the length of time the follow-up would be conducted would vary, depending on the total effective dose. It is recommended that individuals with an estimated total effective dose of <1 mSv y–1 not be actively monitored for medical treatment purposes or followed to identify adverse health effects (ICRP, 2005). The remaining individuals would be those with the lowest (or no measurable) dose, those who were nearby but far enough from the center location of the incident to have little or no likelihood of internal contamination, and individuals who self-reported but had no measurable surface contamination. Neither bioassay sampling nor additional measurements for internal contamination would take place for this group. However, if logistically feasible, these individuals should be registered and followed. Information about these individuals is important in defining baseline rates of disease in the

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population or in the vicinity of the incident, and could be important in establishing, evaluating and assessing claims of illness compensation later on. During the initial phase of the incident, an emergency plan must be quickly implemented to guide the identification and registration of individuals who will be monitored using bioassay methods, and who will be followed for a longer period of time. This will likely be based on geographical information regarding probable exposure levels that were experienced by individuals in a range of locations, from the incident site to neighboring and distant communities. Mapping of how far, how quickly, and in what direction radioactive materials traveled will be very important, as will information on wind, weather patterns, and meteorological conditions at the time of and after the incident. For any significant incidents involving large numbers of people, it will not be feasible to identify all exposed individuals, or to monitor or interview them all on an individual basis. Some people will present to decontamination stations and others will not. For those presenting at a decontamination facility, it will be important to collect baseline information as part of the initial screening process. These data are essential, and may provide information for future epidemiological studies and may assist in making decisions regarding the need for a more extensive follow-up program. It may be also helpful for epidemiological purposes to select samples according to subgroups, for example, according to age, profession, residential area, or extent of local contamination. Within each of these groups, there may also be other “critical” factors that might be useful to distinguish (e.g., specific dietary practices). The number of people sampled from each group or category may necessarily be restricted by the availability of resources or by the extent of cooperation by their members. Nevertheless, care should be taken to include responders in the initial screening, as well as the subsequent follow-up program. It may also be necessary to make special arrangements for pregnant women, young children, and family animals (i.e., pets). The extent to which an individual would be actively monitored and followed to identify the development of adverse health effects, and the length of time such follow-up would continue, should be determined based on the estimated total effective dose for that individual, and the likelihood of health impacts that might occur as a result of that dose. As noted above, the most highly exposed individuals should receive the most attention. They may require medical treatment for internal contamination, and thereafter should be placed under long-term bioassay surveillance.

170 / 11. LONG-TERM FOLLOW-UP OF INDIVIDUALS Long-term surveillance would consist of regular contact with each individual (e.g., urinalysis once or twice a year), and verification of dose and health status (which could consist of a medical examination that would require a visit to a clinic, or simply completing a questionnaire, which could be done at home). Individuals with intermediate level exposure should be monitored in the same manner as those with the highest exposure, and many may need only to be placed under long-term surveillance (i.e., do not need medical treatment). Long-term surveillance in this case would consist of regular contact and the completion of a questionnaire regarding dose and health status. The decision of which people would require more intensive bioassay measurements initially will depend on the route of exposure, radionuclide and form involved as well as the level of intake. Individuals with little or no exposure will not require surveillance or bioassays unless their samples are intended for special study purposes. In that case, depending on the number of individuals registered and the resources available to follow individuals for a long term, it might be possible to identify and actively follow a subgroup of individuals having characteristics representative of the larger group (e.g., age distribution, gender, ethnic group). The specific details of how such a sample would be selected, and the specific objectives of identifying and following a subgroup, would depend on underlying study aims. Whereas medical care of the exposed persons is important, it is also important that efforts be made to minimize social-psychological impacts (as described more fully in Section 10). An important element of such efforts is to verify and validate all bioassay measurements, if possible. Key objectives are to ensure that each bioassay measurement is conducted under the best circumstances possible (Section 8). For example, urine samples should be collected in an area free of contamination from persons’ clothing, and free of contamination from urine sample handling and labeling. All samples must also be kept under stringent chain of custody control (Sun et al., 1993). While collection of additional samples is important in documenting and validating the body contents, such samples are also important in confirming the identities of the radionuclides that are involved, and that the person is not continuing to ingest or inhale various contaminants. Overall, to prevent overreaction or to reduce economic and social-psychological impacts, it is essential that a risk-informed approach be applied (e.g., that protection decisions are based on the magnitude of the likely radiological hazard and risk).

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11.3 Follow-Up Procedures The number of people selected for follow-up and the determination of whether to follow specific individuals and for how long will depend on the type of incident, the potential levels of exposure, and the resources available. Similarly, the specific tests to be administered throughout a follow-up period and the frequency of active follow-up and testing will be specific to the incident. In general, sound implementation plans for radiological emergencies have been developed and are available (NCRP, 1991; 2001; 2005). In addition, detailed written procedures should be developed based on standard epidemiological methods to ensure that follow-up is as complete as possible. Medical tests administered and the results of those tests should be recorded in a standardized manner. Additional data collection instruments may be developed as part of the operating procedures. In terms of interpretation of bioassay results and dose assessments, guidance is described in NCRP (1987; 2008a). Further, it is recommended that any follow-up program employ active rather than passive methods of follow-up. Examples of active follow-up methods include direct contact by mail and telephone. Examples of passive methods include using death records for mortality endpoints and using cancer registry records for cancer incidence. It is also recommended that a comprehensive database be established for any follow-up program that is started which includes the initial data collected at the time of or shortly after the incident, as well as all subsequent information collected on each individual. This database should be maintained securely and only be accessed by those individuals responsible for the program (as specified in the consent form). Individuals who are identified for follow-up should be requested to sign a written consent form that specifies what will be done with the information to be collected and who will have access to it. Once the consent form is endorsed, a questionnaire should be administered to collect basic demographic and contact information, as well as information that will be helpful in estimating exposure level or radiation dose. It will be difficult under trying circumstances, particularly if large numbers of people are identified for follow-up, to collect all the information that may be required for rigorous follow-up. Thus, the amount of information collected initially should be kept to the minimum as much as possible. Although this interview instrument may need to be tailored to the specific incident, a sample registry form is provided in Appendix D.2. Alternatively, if time and resources allow, the ATSDR Rapid Response Registry Survey Form could be used (Appendix D.3). Specific symptoms, radiological findings, and medical laboratory results should be entered into the patient record.

12. Scalability Scalability refers to the change in response capability when the number of affected or potentially affected individuals increases from one or two to hundreds or thousands. Scalability is important to response planning. Activities that can be readily undertaken for one or two individuals can be completely impractical when dealing with large numbers. The medical community applies principles of scalability in mass casualty triage when resources are limited at first but are expanded to meet the increasing demand of the emergency (Burkle, 2006; Devereaux et al., 2008). Similar logic must be applied to screening and monitoring of the public following the spread of contamination from a large RDD or IND. Screening for internal contamination and standard medical procedures probably will need to be modified to reflect actual conditions and availability of resources (DHHS, 2009). This section provides guidance on planning and allocating limited resources for response to public contamination incidents. There are no firm boundaries for what constitutes small versus large in scalability as it applies to emergency response. Each responding facility or planning entity must make its own determination based on its role and on its ability to expand available resources in response to an emergency (DHHS, 2009). The guidance provided here is divided into three broad classes, a small-sized group consisting of 1 to 10 people, a medium-sized group (tens of people), and a large-sized group (over 100 people). It is readily recognized that for a small private or community facility, the grouping may be too large, and tens of people could immediately overwhelm available resources. The group classifications are provided for relative comparison of actions only. The actions associated with screening a population for internal contamination may include predecontamination surveys of subjects, personal decontamination actions, post-decontamination surveys of subjects, a history questionnaire, urine sampling as radiobioassay measurements (if necessary), direct (in vivo) radiobioassay measurements, dose-reduction therapy, dose assessment, and long-term follow-up. In the event of large-scale contamination from a radiological or nuclear incident, resources to accomplish these tasks will need to increase directly with the size of the incident and the number of people who need to be monitored (DHHS, 2009; HSC, 2009). 172

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As discussed in Sections 6 and 8, if the number of people potentiallycontaminated internally exceeds available screening resources, a strategy may need to be developed to identify groups of people who are more likely to have high levels of internal contamination. Medical resources for treatment of acute injuries will need to increase directly with the number of injuries identified in triage. Medical follow-up (i.e., complete blood counts and other laboratory tests,) and treatment of persons with radionuclide intakes of >1 CDG are less likely to be severely impacted because the number of patients expected to have intakes in excess of the CDG is not expected to be large even if thousands of people need to be monitored (DHHS, 2009). However, a small hospital laboratory could become overwhelmed if laboratory tests are not ordered judiciously. The Goiânia, Brazil incident demonstrates the importance of scaling-up the screening, monitoring, treating, and following-up of the population (i.e., monitoring 112,000 people who may have been exposed to the radiation source). The recommendations for scaling based on size of group are provided in Table 12.1 for each of these activities. Table 12.2 provides an example of emergency department scalability with three major response categories and multiple response tiers based on scope and number of casualties. In the event of a nuclear detonation, priority care should be given to burn and blast injuries and ARS effects (HSC, 2009). Treatment for internal deposition of radionuclides is a significantly lower priority and should not divert limited medical resources from burn and blast care. When practicable, the procedures outlined in this Report can be used for establishing long-term follow-up and therapy. In conclusion, community and hospital radiation emergency plans should be scalable. Entities that have limited resources should build into their plans connections to larger entities that could help out when needed. Such connections can be accommodated in regional or state radiation emergency plans and identified through memoranda of agreements among the entities within a city or region. Scalability should be considered in each community or hospital radiation-response plan.

Hours to 1 d Single facility (e.g., hospital). Not likely required.

Survey each patient for external contamination. Within a few hours after the incident, collect nasal smears if inhalation exposure was possible.

Facility needed

Security needs

Predecontamination survey

Small Group (1 – 10)

Time for effective response

Activity or Task

Optional depending on resources or go directly to decontamination without the survey.

Ask for assistance from facility security personnel.

Multiple facilities (e.g., two or more hospitals).

1–2d

Medium Group (tens of people)

Go directly to decontamination without survey.

May need to supplement facility security.

Screening center, possibly a regional sports stadium.

2+ d

Large Group or Population (>100)

Population (thousands)

Go directly to decontamination without survey.

Utilize National Guard for security at hospitals and health assessment centers.

One or more large facilities for screening (stadium).

Several days to weeks

TABLE 12.1—Suggestions for scaling of response to a radiation incident.

174 / 12. SCALABILITY

Supervised decontamination of each patient. Survey after decontamination or as process continues.

Staff takes information or patient fills out own form.

Decontamination

Post-decontamination survey

History questionnaire

Staff takes information or patients fill out their own form.

Survey after decontamination.

Decontamination by shower or supervised decontamination of each patient.

Patients fill out their own form.

Survey only after verifying decontamination has been performed.

Large scale shower decontamination or have patient wash/ shower at home.

Patients fill out their own form. Request administrative support to process.

Survey only after verifying decontamination has been performed. Request assistance from state or federal government.

Large scale shower decontamination or have patient wash/ shower at home.

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Collect single-void sample as backup for screening. Screen with survey meter. Obtain 24 h collection. Send to appropriate laboratory for analysis.

Screen each patient as appropriate. Use GM survey meter if contaminant is a high-energy photon emitter.

Direct (in vivo) screening measurement

Small Group (1 – 10)

Urine sample

Activity or Task

Screen each patient as resources allow. Select representative patients for measurement based on limited resources.

Collect single-void sample (nominal 100 mL volume) for screening. Screen individually with survey meter. Collect 24 h sample from each or representatives of group as resources allow. Send to appropriate laboratory for analysis.

Medium Group (tens of people)

TABLE 12.1—(continued)

Probably not a feasible early response; screen injured patients only. May need time (many hours to days) to obtain and organize resources to screen all.

Collect single-void sample (nominal 100 mL volume). Screen individually with survey meter if resources allow. Send to appropriate laboratory for analysis.

Large Group or Population (>100)

Not feasible as part of early response. Screen injured patients only. Will need days to screen all.

Collect single-void sample (nominal 100 mL volume) only from patients who have positive screening survey. Screen samples individually with survey meter if resources allow. Send to appropriate laboratory for analysis.

Population (thousands)

176 / 12. SCALABILITY

As appropriate based on patient presentation and available data. Individual-specific assessment, presumably by knowledgeable dosimetrist. Relatively easy to decide. Identify responsible party.

Dose-reduction therapy

Dose assessment

Long-term follow-up

Relatively easy to decide. Identify responsible party.

Probably individual dose assessments by oversight or responsible agency. May be individualized or broad categories.

Decide based on patient presentation and available data to the extent resources allow.

Complex. Will likely need coordination by overseeing agency.

Conducted by oversight or responsible agency. May involve broad categories for dose assignment.

Triage based on simple indicators such as patients with injuries who failed screening.

Complex. Will likely need coordination by overseeing agency.

Conducted by oversight or responsible agency. May involve broad categories for dose assignment.

Triage based on simple indicators such as patients with injuries who failed screening.

12. SCALABILITY

/ 177

Emergency department only

1 – 4 adult 1 – 2 pediatric 1 – 10 None

Red triage casualtiesb

Total patients

Radiological incident

Tier 1

Tier

Scope

Normal Operations

Level

Numerous blast injuries 3 – 20 patients contaminated externallyc

1 – 2 patients contaminated

21 – 40

11 – 20 adult 6 – 10 pediatric

Hospital wide

Tier 3

Small incident; few patient blast injuries

11 – 20

5 – 10 adult 3 – 5 pediatric

Assistance from radiology, surgery and radiation safety

Tier 2

Multiple Casualties

20 – 50 patients contaminated externally

Numerous blast injuries

41 – 100

>20 adult >10 pediatric

Community wide

Tier 4

51 – 100 patients contaminated externally

Numerous blast injuries

101 – 500

>20 adult >10 pediatric

Regional

Tier 5

Mass Casualties

>100 patients contaminated externally

Numerous blast injuries

>500

>20 adult >10 pediatric

National

Tier 6

TABLE 12.2—Example of mass casualty incident response tiers for emergency department response to injuries from an explosive radiological device.a

178 / 12. SCALABILITY

Not implemented

Limited

Expanded

Full

9Sztajnkrycer,

9

Full/unified command

11 – 20 patients contaminated internally

M. (2009). Personal communication (Mayo Clinic, Rochester, Minnesota).

Modeled after Mayo Clinic Level 1 Trauma Center Mass Casualty Incident Response.9 Triage category “Immediate”; life threatening injury. c Assumes that most patients were decontaminated before arrival at emergency department. d Hospital Incident Command System (Section 5.3).

b

a

HICSd

1 – 10 patients contaminated internally Unified command

>20 patients contaminated internally

12. SCALABILITY

/ 179

13. Assessment of Current Capacity in the United States to Perform Population Screening, Decontamination and Monitoring for Internal Contamination Section 8 described an approach to screen a potentially-contaminated population for internal contamination using radiation survey meters and existing nuclear medicine equipment. This section presents a snapshot of past and current capabilities in the United States to use existing nuclear medicine equipment to perform assessments of internal radionuclide contamination of patients during a mass casualty incident. The information presented in Section 13.1 was compiled from three sources: • a survey of gamma cameras conducted by Phillips Medical Systems in 2000;10,11 • a survey of state radiation-control programs conducted by CRCPD to determine the availability of equipment in hospitals that could be used to screen or monitor patients for internally-deposited radionuclides following exposure during an RDD or IND incident;12 and • a quarterly survey of critical assets available at New York State hospitals, conducted by the New York State Department of Health that included questions on availability of radiation-detection equipment.12 10Garrard,

J.L. (2000). Personal communication (Phillips Medical Systems, Andover, Massachusetts). 11Wood, C.M. (2007). Personal communication (Centers for Disease Control and Prevention, Atlanta). 12Salame-Alfie, A. (2008). Personal communication (New York State Department of Health, Troy, New York).

180

13.1 AVAILABILITY OF EQUIPMENT AND RESOURCES

/ 181

13.1 Availability of Equipment and Resources Even though the type and number of gamma cameras available at hospitals differ at various locations across the country mainly due to advances in technology and replacement of old equipment, gamma cameras are a resource that should be factored into community and hospital radiological emergency response plans. There is large variability in the types of gamma cameras available; the survey in 2000 of gamma cameras available at hospitals included 21 listed manufacturers and over 580 models installed. There were many different factors specific to individual cameras such as the thickness of the crystal (ranging from three-eighths to one inch), types of collimators (low, medium and high energy, with low energy being used for most of the routine diagnostic studies), and energy windows. Due to this variability, it was recommended that hospitals consider contacting the manufacturers to request specific modifications to the computer programs for their gamma camera to include a preset on the camera to allow for the detection of the gammaemitting radionuclides that may be found after an RDD or IND attack. Furthermore, it was also recommended that hospitals be requested to develop specific calibration factors for their cameras to minimize the need for the operators to manually perform those changes, thus allowing them to quickly convert their cameras to this nontraditional use. Results from a survey of RCPs13 showed that the percentage of hospitals with gamma cameras among the responding states range from 5 to 100 % and the availability of mobile cameras range from 0 to 46 %. The availability of thyroid counters at hospitals ranges from 6 to 75 %, though some radiation-control program responders indicated that hospitals could use the gamma cameras to measure thyroid uptakes of radioiodine. All nuclear medicine departments are required by regulation to possess calibrated radiation detection and measuring instruments, but it is difficult to assess the availability of GM survey meters outside of emergency departments. To obtain information on availability of survey meters at hospitals, it is necessary to survey the hospitals. Results from a survey of critical assets available in New York State hospitals provided an indication that such equipment is readily available at 82 % of the hospitals.13 The survey also provided data on the percent of hospitals that have pocket dosimeters (58 %) and the percent of hospitals that have portable radiation detectors 13Salame-Alfie,

A. (2008). Personal communication (New York State Department of Health, Troy, New York).

182 / 13. ASSESSMENT OF CURRENT CAPACITY available in the emergency department (11 %). Other information that was gathered from the New York State hospital survey included the percentage of hospitals that have a decontamination team (76 %), and, of those, 9 % have a separate decontamination plan for pediatric and special-needs patients. The majority of hospitals do not have portable portal monitors. Though many hospitals (~50 %) have waste-monitoring stations, those monitors are fixed and located at the waste-handling area. It may be possible to use them during an emergency if contaminated victims are directed to enter the hospital via the waste-handling area or if the portal monitor can be moved to the emergency department. About 68 % of the RCPs that responded to the survey have purchased portable portal monitors or have made arrangements to use portal monitors available at nuclear power plants or national laboratories.14 Most states that have nuclear power plants have access to whole-body counters (mostly located at the nuclear power plants). Few states have mobile decontamination facilities or equipment. Most RCPs reported that there is a National Guard civil support team in their state, but that the RCPs do not share resources with the civil support teams. Most state programs have access to spectrum analyzers within their program. Only a few state programs keep a list of physicists, medical physicists, dosimetrists, and radiation-safety officers that may be available to assist during a radiological emergency, though some have identified volunteers within these groups and have a program setup to qualify them as volunteers. 13.2 Laboratory Capabilities A critical component of a bioassay program is the ability to analyze the biological samples collected. Most hospitals are able to collect urine and blood samples during routine diagnosis. Traditionally, bioassays to assess internal radionuclide depositions in occupational settings are conducted on 24 h urine samples. It has been recognized that this protocol would not work during a radiological mass casualty incident due to the number of samples and the difficulty of obtaining and analyzing such a large number of samples. To address this issue, CDC laboratories have been working on a new methodology to analyze “spot” urine samples that involve a small amount of urine. As discussed in Section 8, CDC is also in the process of obtaining Clinical Laboratory Improvement 14Salame-Alfie,

A. (2008). Personal communication (New York State Department of Health, Troy, New York).

13.3 TRAINING NEEDS ON USE OF EQUIPMENT

/ 183

Amendments approval for a gross-alpha/gross-beta method using liquid scintillation and is developing a method to use gamma spectroscopy to conduct quick screening of urine samples to expedite analysis of those that require detailed isotopic analysis. General information on the collection and shipping information is provided in Appendices F and G. Most hospitals are traditionally able to collect and analyze blood samples for complete blood cell counts with differentials. The results of these analyses will provide doctors with an initial indication of high radiation exposure from an external source. CDC is also working on new protocols for blood sample analyses.15 As discussed in Section 8, the use of fecal samples to determine internal contamination during a radiological incident with mass casualties is not practical. These samples have often been collected to assess occupational exposures. The logistics of sample collection and complex analytical processes render then unsuitable for mass casualty applications. In addition, only a few laboratories have analytical capabilities to analyze them and mass casualty scenarios would overwhelm their capacities. While there are a number of laboratories that can analyze environmental samples for activity, very few laboratories in the country have the capability and capacity to conduct bioassays. Results of the CRCPD survey16 indicated that ~72 % of the existing state laboratories are able to handle radiological environmental samples in-house, but very few could do radiological clinical analyses. Some states have identified commercial laboratories that can handle such samples, but few have an existing contract/agreement to perform these analyses. 13.3 Training Needs on Use of Equipment To use the equipment described above, hospital staff need to be trained in its proper use. Every hospital that has a nuclear medicine department has at least one staff person trained in the use of the gamma camera and survey meters. During a mass casualty incident, the nuclear medicine staff will not be able to support the needs for quick screening of patients and they will need other hospital staff available to assist with these activities. Hospitals should have staff trained in areas other than their traditional roles, such as in the use 15Jones, R. (2008). Personal communication (Centers for Disease Control and Prevention, Atlanta, Georgia). 16Salame-Alfie, A. (2008). Personal communication (New York State Health Department, Troy, New York).

184 / 13. ASSESSMENT OF CURRENT CAPACITY of survey meters, portal monitors, contamination control, sample labeling, conducting screening surveys, decontamination, etc. Based on the activities that are expected to take place at the hospital in response to a radiological mass casualty incident, it is recommended that hospital staff be trained in the following areas: • radiation safety and the effects of radiation on patients; • effects of radiation and radioactive contamination on a developing fetus; • how to determine whether a victim is contaminated; • collection and measurement of nasal swabs; • how to survey wounds for evidence of radioactive shrapnel or for radionuclide contamination; • wound cleaning and treatment if radionuclide contamination is detected or suspected; • how to determine the severity of radiation injury, the potential for radiation injury to worsen, the impact of nonradiation injuries on radiation injury, and other risks that may be involved; • current strategies for surveying large numbers of people (including walk-through whole-body monitors and rapid personnel surveys using hand-held instruments); • decontamination of large numbers of people (mass decontamination); • performing radiological surveys (including identifying alpha, beta or gamma radiation): - interpretation of the survey meter readings; - how to perform a contamination survey; and - contamination-control practices: assessing the need for contamination-control measures, and contamination-control of the victim. • contamination-control among medical and emergency response personnel; • use of personal protective equipment; • examples of contamination-control measures: - ambulance and treatment area contamination control; - contamination-control actions in the emergency department; - working with contaminated patients; - instructions for leaving a controlled area: patient and responder; and - decontamination (self, patient and equipment). • record-keeping; • patient records;

13.4 RADIATION VOLUNTEERS

/ 185

• workers (emergency responders); and • proper labeling of laboratory samples (includes using specific containers/documentation with the samples). 13.4 Radiation Volunteers to Support Population Screening The Radiation Studies Branch of CDC supports and has been working with several state radiation-control programs to establish a Radiation Volunteers Corps, similar to the Medical Volunteers Corps. The Florida Bureau of Radiation Control has established a radiation volunteer program and the information presented here is based on their program.17 The initial steps in the development of this program included the determination of: • • • •

intended duties of the corps; professions that may qualify with limited training; whether a mechanism for volunteering already exists; infrastructure needed for staging a population screening or community reception center; • providing initial training; and • supporting annual training, infrastructure, and logistics. The radiation response volunteers thus identified will not be considered emergency responders but will help fill the gap identified in the National Response Framework, Radiation Response Annex section dealing with population screening. These individuals are expected to be able to help in response to the need to monitor large numbers of people at the community reception centers, population screening centers, entrance to Red Cross shelters, or other locations identified in the emergency response plan for this activity. These individuals are also expected to “staff-up” between 12 h after the incident and until such time as federal assets can be mobilized. The volunteers are individuals trained in identification of contamination and implementation of decontamination procedures as part of their current job duties and may have some of the following additional qualifications: • have knowledge and are experienced in answering questions regarding health risk from radiation; • are able to collect or use epidemiological information; 17Gilley,

D.B. (2009). Personal communication (State of Florida Department of Health, Bureau of Radiation Control, Tallahassee).

186 / 13. ASSESSMENT OF CURRENT CAPACITY • have training and experience in disaster mental health; and • can provide “reach-back” supervision directly with the radiation-control program through established communication channels. The Florida program determined that there was already an existing mechanism for volunteering through the Medical Reserve Corps, which is a specialized component of the Citizen Corps. This group has medical and public-health professionals ready to serve their communities in times of need and more information on this resource can be found on the Medical Reserve Corps website (MRC, 2011). Other states may want to consider establishing a similar mechanism for a volunteer corps. 13.5 Biodosimetry The proceedings of the BioDose 2008 International Symposium (Simon et al., 2010) are an excellent resource on the current status and future direction of biodosimetry for high-level radiation exposures. In these proceedings, Swartz et al. (2010) provide a critical assessment of biodosimetry methods for large-scale incidents. They focus on the use of biodosimetry to: • identify individuals who did not receive a significant radiation exposure; • classify the exposed persons into different treatment categories as needed; and • guide both short- and long-term medical treatments. Knowledge of the strengths and weaknesses of various biodosimetry techniques related to these three stages will be critical to their effective use in managing a mass-casualty event. One new approach for early-response assessment of radiation exposure that shows promise is measurement of plasma protein serum amyloid A as a complementary approach to conventional biodosimetry for early assessment of radiation exposures and, when coupled with peripheral blood cell counts, provides early diagnostic information for the effective management of radiation casualty incidents (Ossetrova et al., 2010). However, current national resources for biodosimetry are limited and need to be enhanced to provide timely and adequate dose assessments (Blakely et al., 2005). Physicians who are considering these biodosimetry methods should contact REAC/TS (Appendix K) for assistance. High throughput biodosimetry will be needed if large numbers of people are exposed to an IND and possibly an RDD. This is well illustrated by the 1987 radiation incident in Goiânia, Brazil

13.6 CONCLUSIONS

/ 187

(Appendix I) where ~112,000 people reported for screening, of whom 46 required treatment with Prussian blue (IAEA, 1988). Identifying those patients who do not need medical intervention will be equally crucial to reduce demand on limited resources and to reassure those patients. New, fully-automated approaches to biodosimetry have the potential for much higher throughputs (Garty et al., 2010). The World Health Organization has formulated the general scope and concept of a global biodosimetry laboratory network for radiation emergencies (BioDoseNet) and is developing the technical details of the network (Blakely et al., 2009). 13.6 Conclusions The capability to use nuclear medicine equipment available at hospitals and clinics exists in every state. Many operators, however, are not trained in how to change the settings to allow them to use the gamma cameras to measure internal contamination resulting from a radiological incident. It is unknown at this time which hospitals have developed protocols to allow them to use the equipment during a mass casualty radiological incident. Therefore, if a hospital is planning to use this equipment during a mass casualty incident, it is in their best interest to request modifications to the gamma camera’s computer from their vendors, and to request that they develop specific calibration factors to minimize the need for the operators to manually perform those changes. Hospitals also need to develop hospital-specific protocols for handling contaminated patients to minimize contamination of equipment and other areas of the hospital. Staff using the equipment should be trained in their use during an emergency, and this should be made an integral part of the plans developed by the hospitals to quickly assess contaminated patients. Additionally, using the equipment in this fashion should be practiced during hospital emergency exercises. Finally, additional research should be conducted to determine whether concerns of emergency personnel to report to a radiological or nuclear incident will impact the capacity of agencies to respond adequately. Hospitals should be aware of other statewide resources available that may be used to supplement their resources. This could be accomplished by contacting their state radiation-control program. A list of contacts for these radiation-control programs is available from CRCPD (2010).

14. Conclusions and Recommendations This Report has discussed screening a population for internal contamination following exposure to radionuclides that are considered most likely to be involved in a radiological or nuclear incident. The Report contains many recommendations that focus on procedures for screening a population to determine whether individuals have radionuclide intakes that exceed the CDG. The CDG was developed in NCRP Report No. 161 (NCRP, 2008a) as an operational quantity that physicians could use as a basis for medical treatment of individuals having one or more internally-deposited radionuclides. Many communities and hospitals are likely to be capable of screening a few tens of exposed individuals, but little information exists to determine the capability of screening large numbers of people who may have been exposed to the radionuclide(s) involved in a radiological or nuclear incident. 14.1 Recommendations for Planning The following recommendations are highlighted in the Report and are intended to help local communities plan and prepare to screen a population for internal contamination following a radiological or nuclear incident. The information provided in this Report should be used to evaluate and to upgrade, to the extent feasible, the capability of a local community to screen small, medium and large populations for the presence of internally-deposited radionuclides and to treat members of the population medically to reduce radiation dose to patients who contain high levels of one or more radionuclides based on comparisons of these internal depositions with CDG values in this Report or other guidelines. • In the development of a response plan for a radiological or nuclear incident, community leadership should identify personnel such as health physicists, radiation-safety officers, or medical personnel who could serve as subject matter experts or perform specific tasks to support response to a radiological or nuclear incident. • Screening a population for internal contamination requires planning and practice. In particular, the use of nuclear 188

14.2 RECOMMENDATIONS RELATED TO SCREENING

/ 189

medicine cameras to screen patients following exposure to radionuclide contamination from an RDD or IND incident requires advance planning, training and rehearsal. It cannot be implemented ad hoc during an emergency. • Local municipalities and public-health agencies must develop procedures for requesting, receiving and distributing SNS assets. This planning should include providing the diagnostic and medical management guidelines to healthcare providers for the use of decorporation agents for internally-deposited radionuclides. Lists of possible decorporation agents for radionuclides beyond those covered in this Report are given in Sections 3 and 12 of NCRP Report No. 161 (NCRP, 2008a). 14.2 Recommendations Related to Screening and Treatment of a Population for Internal Contamination The following points are emphasized to help local communities screen (and treat, if necessary) members of a population who may have been internally contaminated during a radiological or nuclear incident such as an RDD or IND: • Patients who have suffered life-threatening injuries should be given medical care immediately, without regard to contamination. • People with the most serious injuries are also likely to be the most contaminated both externally and internally. • The presence of internal contamination is rarely life-threatening and decorporation should not take precedence over treatment of conventional injuries that may be acutely life threatening. • The CDG may be used by physicians as a basis for considering medical treatment of individuals who have an internal radionuclide deposition. • The CDG is not intended to instruct physicians on a specific course of action such as administration of decorporation agents. Rather, the CDGs are intended as a tool to be used to help a physician determine when radiation exposure may have clinical significance. • REAC/TS should be used as a resource for information and assistance when considering the treatment of internallycontaminated individuals (REAC/TS, 2010).

190 / 14. CONCLUSIONS AND RECOMMENDATIONS • Rapid identification of the radionuclide(s) involved in a radiological or nuclear incident is essential to the selection of appropriate methods for assessment of internal contamination and subsequent treatment decisions. • The rapid detection and identification of pure alpha- or betaemitting radionuclides (e.g., 210Po, 90Sr/90Y, respectively) are challenging and should be addressed in the planning process. • In all but the most extreme cases, standard precautions provide adequate protection to healthcare workers to prevent secondary contamination. The presence of external contamination should almost never delay urgent medical care. • The use of nuclear medicine cameras to screen patients following exposure to radionuclide contamination from an RDD or IND incident requires advance planning, training and rehearsal; it cannot be implemented ad hoc in an emergency. • Local municipalities and public-health agencies must develop procedures for requesting, receiving and distributing SNS assets. This planning should include providing the diagnostic and medical management guidelines to healthcare providers for the use of decorporation agents for internally-deposited radionuclides. • Planning for population screening should incorporate provisions for establishing stakeholder advisory boards or other mechanisms for making the public a partner. • NCRP recommends a bioassay action-level or benchmark equal to one-half of the CDG value for identifying those individuals for whom continued bioassays should be considered for the purpose of correlating internal contamination with long-term biological effects. • Scalability should be considered in each hospital or community radiation-response plan. 14.3 Recommendations for Additional Work The following recommendations should be implemented by appropriate federal agencies to improve capability of the nation to screen populations for internal contamination following exposure to a radiological or nuclear incident: • Additional surveys of local communities and states should be conducted to determine more accurately their capability to screen populations for internal contamination. • Consideration should be given to utilizing strategicallylocated radiochemistry laboratories to analyze bioassay samples for radionuclide identification and concentrations. A

14.3 RECOMMENDATIONS FOR ADDITIONAL WORK

/ 191

limiting factor here is the low number of available radiochemists in the United States. • Additional research should be conducted to define radiological instrument or bioassay trigger levels that correspond to the CDG for a wider variety of radionuclides such as those identified in NCRP Report No. 161 (NCRP, 2008a) that may be produced by the explosion of an RDD or IND.

Appendix A Radiological Properties of Radionuclides Considered in this Report Tables A.1a and A.1b are a compilation of the radionuclides considered in this Report and their relevant radiological data in SI and previous units, respectively, as given in NCRP Report No. 161 (NCRP, 2008a) and drawn in large part from ICRP Publication 107 (ICRP, 2008). These data include half-life, decay mode, energy, and specific activity of the primary radiation. Also given are air kerma rate constants and electron constants. The air kerma rate constants are radiation doses in air at a distance of 1 m from an unshielded point source of unit activity [1 Bq (1 μCi)]. The electron constants are radiation doses at a depth of 70 μm from a 1 cm2 source of unit activity [1 Bq (1 μCi)] on the skin surface. For those radionuclides that typically exist as equilibrium mixtures with their progeny (e.g., 137Cs/137mBa and 90Sr/90Y), the air kerma rate and electron constants are provided for both the parent and progeny radionuclides; progeny radionuclides are indicated with an asterisk. The total radiation dose from an equilibrium mixture of a parent radionuclide and its progeny is the sum of the doses from both the parent and progeny. Tables A.1a and A.1b provide dose rate information from radionuclide sources external to the body (at 1 m and on the skin surface).

192

Co

8.02 d

11.8 d 30.2 y 2.55 m 73.8 d

I

*131mXe

Cs

131

137

*137mBa

192

226Ra

See 238U chain

64.1 h

*90Y

Ir

28.8 y

5.27 y

Half-life

Sr

90

60

Radionuclidea

4.6 × 106 3.1 × 106 3.2 × 103 2.0 × 1010

β 2.28 γ 0.364, β 0.807 γ 0.0298 β 1.176 γ 0.662 γ 0.316, β 1.38

β– β–

IT β– IT β– EC

3.4 × 105

2.0 × 107

5.1 × 103

β 0.546

β–

4.1 × 104

Specific Activity (GBq g–1)

γ 1.17, 1.33 β 2.51

Energy of Prominent Radiations (MeV)

β–

Decay Modeb

3.2 × 10–17

2.3 × 10–17

6.1 × 10–23

3.3 × 10–18

1.5 × 10–17

1.7 × 10–21

—

8.5 × 10–17

Air-Kerma Rate Constant [Gy s–1 (Bq m–2)–1]

5.5 × 10–10

6.9 × 10–11

4.7 × 10–10

6.5 × 10–10

4.8 × 10–10

6.7 × 10–10

5.0 × 10–10

3.1 × 10–10

Electron Constante [rad h–1 (µCi m–2)–1]

TABLE A.1a—Radiological properties of radionuclides in this Report given in SI units (NCRP, 2008a).

A. RADIOLOGICAL PROPERTIES OF RADIONUCLIDES

/ 193

γ 0.0633, β 0.195 γ 1.00, β 2.27 γ 0.0136, β 2.072

β– β– IT β– α

24.10 d

1.17 m

6.70 h 2.5 × 105 y

*234Th

*234mPa

*234Pa

*234U

γ 0.0163, α 4.78

γ 0.016, α 4.20

α SF

U

4.5 × 109 y

β 0.389

238

γ 0.186, α 4.397

α β–

7.1 × 108 y

Energy of Prominent Radiations (MeV)

25.5 h

U

Half-life

Decay Modeb

*231Th

235

Radionuclide

a

2.3 × 10–1

7.4 × 107

2.5 × 1010

8.6 × 105

1.2 × 10–5

2.0 × 107

8.0 × 10–5

Specific Activity (GBq g–1)

TABLE A.1a—(continued)

2.8 × 10–18

7.4 × 10–17

7.4 × 10–19

2.9 × 10–18

2.0 × 10–18

1.9 × 10–17

1.3 × 10–17

Air-Kerma Rate Constant [Gy s–1 (Bq m–2)–1]

9.3 × 10–13

9.2 × 10–10

6.6 × 10–10

1.1 × 10–10

2.7 × 10–13

2.5 × 10–10

4.7 × 10–11

Electron Constante [rad h–1 (µCi m–2)–1]

194 / APPENDIX A

γ 0.609, β 3.27 α 7.69 γ 0.800, β 4.394 γ 0.108, β 0.0631

β– α α β– β– α

19.9 m 1.6 × 10–4 s 1.30 m

22.2 y

*214Po

*210Tl

*210Pb

*214Bi

β 2.88, α 6.69

α β–

1.5 s

*218At

γ 0.352, β 1.02

β–

26.8 m

*214Pb

β 0.259, α 6.00

α β–

3.10 m

*218Po

α 5.49

α

3.82 d

*222Rn

γ 0.186, α 4.78

α

1,600 y

*226Ra

γ 0.0153, α 4.69

α

7.5 × 104 y

*230Th

2.8 × 103

2.5 × 1010

1.2 × 1016

1.6 × 109

1.3 × 1012

1.2 × 109

1.0 × 1010

5.7 × 106

3.7 × 101

7.6 × 10–1

9.7 × 10–18

1.0 × 10–16

3.1 × 10–21

5.0 × 10–17

—

1.5 × 10–17

—

1.5 × 10–20

5.2 × 10–19

2.5 × 10–18

—

8.4 × 10–10

7.8 × 10–16

6.4 × 10–10

6.8 × 10–13

6.9 × 10–10

4.2 × 10–14

1.6 × 10–14

1.4 × 10–11

1.8 × 10–12

A. RADIOLOGICAL PROPERTIES OF RADIONUCLIDES

/ 195

432 y

241

Am

26 m

*235mU IC 0.000076 γ 0.0600, α 5.49

IT α

γ 0.0136, α 5.176

α

2.4 × 104 y

Pu

239

γ 0.0173, α 5.46, α 5.50

α SF

87.7 y

Pu

238

β 1.53

β–

4.2 m

*206Tl

γ 0.305, β 1.31

8.15 m

*206Hg

β–

138 d

*210Po

α 5.30

5.013 d α

Energy of Prominent Radiations (MeV)

β 1.162

Decay Modeb

β– α

Half-life

*210Bi

Radionuclide

a

9.8 × 10–18

—

1.1 × 109 1.3 × 102

1.1 × 10–18

2.6 × 10–18

1.7 × 10–20

6.2 × 10–18

3.6 × 10–22

1.5 × 10–23

Air-Kerma Rate Constant [Gy s–1 (Bq m–2)–1]

2.3 × 100

6.3 × 102

8.0 × 109

4.1 × 109

1.7 × 105

4.6 × 106

Specific Activity (GBq g–1)

TABLE A.1a—(continued)

7.6 × 10–13

—

1.3 × 10–13

2.5 × 10–13

6.2 × 10–10

6.7 × 10–10

8.5 × 10–17

5.9 × 10–10

Electron Constante [rad h–1 (µCi m–2)–1]

196 / APPENDIX A

b

Names preceded by an asterisk are radioactive progeny that may be present in significant quantities. EC = electron capture IT = isometric transition SF = spontaneous fission c Beta energy is the endpoint energy of the spectrum, IC denotes internal conversion electrons, neutrons accompany spontaneous fission. d Includes the contribution from annihilation photons and, in the case of spontaneous fission, the prompt and delayed photons e In the case of spontaneous fission, includes the contribution of prompt beta decay.

a

A. RADIOLOGICAL PROPERTIES OF RADIONUCLIDES

/ 197

Co

64.1 h 8.02 d

11.8 d 30.2 y 2.55 m 73.8 d

*90Y

I

*131mXe

Cs

Ra

137

*137mBa

Ir

131

192

226

See 238U chain

28.8 y

5.27 y

Half-life

Sr

90

60

Radionuclide

a

1.2 × 105 8.4 × 104 8.6 × 101 5.4 × 108

β 2.28 γ 0.364, β 0.807 γ 0.0298 β 1.18 γ 0.662 γ 0.316, β 1.38

β– β–

IT β– IT β– EC

9.2 × 103

5.4 × 105

1.4 × 102

β 0.546

β–

1.1 × 103

Specific Activity (Ci g–1)

γ 1.17, 1.33 β 2.51

Energy Primary Radiationsc (MeV)

β–

Decay Mode

b

4.2 × 10–7

3.0 × 10–7

8.1 × 10–13

4.4 × 10–8

2.0 × 10–7

2.2 × 10–11

—

1.1 × 10–6

Air-Kerma Rate Constantd [Gy s–1 (Bq cm–2)–1]

7.3 × 100

9.2 × 10–1

6.3 × 100

8.7 × 100

6.4 × 100

8.9 × 100

6.6 × 100

4.1 × 100

Electron Constante [rad h–1(µCi cm–2)–1]

TABLE A.1b—Radiological properties of radionuclides in this Report given in previous units (NCRP, 2008a).

198 / APPENDIX A

γ 0.163, α 4.21 γ 0.0633, β 0.195 γ 1.00, β 2.27 γ 0.0136, β 2.07

α SF β– β– IT

γ 0.0153, α 4.687 γ 0.186, α 4.78 α 5.49

α α α α

1.17 m

6.70 h 2.5 × 105 7.5 × 104 y

1,600 y

3.82 d

*234mPa

*234Pa

*234U

*230Th

*226Ra

*222Rn

γ 0.0163, α 4.78

β–

24.10 d

*234Th

U

4.5 × 109 y

β 0.389

238

γ 0.186, α 4.397

α β–

7.0 × 108 y

25.5 h

U

*231Th

235

1.5 × 105

9.9 × 10–1

2.1 × 10–2

6.2 × 10–3

2.0 × 106

6.9 × 108

2.3 × 104

3.4 × 10–7

5.3 × 105

2.2 × 10–6

2.0 × 10–10

7.0 × 10–9

3.3 × 10–8

3.7 × 10–8

9.9 × 10–7

9.8 × 10–9

3.8 × 10–8

2.7 × 10–8

2.5 × 10–7

1.8 × 10–7

2.1 × 10–4

1.9 × 10–1

2.5 × 10–2

1.2 × 10–2

1.2 × 101

8.7 × 100

1.4 × 100

3.6 × 10–3

3.3 × 100

6.2 × 10–1

A. RADIOLOGICAL PROPERTIES OF RADIONUCLIDES

/ 199

γ 0.609, β 3.272 α 7.69 γ 0.800, β 4.39 γ 0.108, β 0.0631

β– α α β– β– α

19.9 m 1.6 × 10–4 s 1.30 m

22.2 y

*214Po

*210Tl

*210Pb

*214Bi

β 2.88, α 6.69

1.5 s

*218At

α β–

26.8 m

*214Pb

γ 0.352, β 1.023

3.10 m

*218Po β–

Decay Mode

Energy Primary Radiationsc (MeV)

β 0.259, α 6.00

Half-life

b

α β–

Radionuclide

a

7.6 × 101

6.9 × 108

3.2 × 1014

4.4 × 107

3.5 × 1010

3.3 × 107

2.8 × 108

Specific Activity (Ci g–1)

TABLE A.1b—(continued)

1.3 × 10–7

1.4 × 10–6

4.1 × 10–11

6.7 × 10–7

—

2.0 × 10–7

—

Air-Kerma Rate Constantd [Gy s–1 (Bq cm–2)–1]

—

1.1 × 101

1.0 × 10–5

8.5 × 100

9.0 × 10–3

9.2 × 100

5.6 × 10–4

Electron Constante [rad h–1(µCi cm–2)–1]

200 / APPENDIX A

432 y

*241Am

IC 0.000076 γ 0.0600, α 5.49

IT α

1.3 × 10–7

—

3.1 × 107 3.4 × 100

1.5 × 10–8

3.5 × 10–8

2.3 × 10–10

8.2 × 10–8

4.8 × 10–12

2.0 × 10–13

6.2 × 10–2

1.7 × 101

2.2 × 108

1.1 × 108

4.5 × 103

1.2 × 105

1.0 × 10–2

—

1.7 × 10–3

3.3 × 10–3

8.3 × 100

9.0 × 100

1.1 × 10–6

7.8 × 100

b

Names preceded by an asterisk are radioactive progeny that may be present in significant quantities EC = electron capture IT = isometric transition SF = spontaneous fission c Beta energy is the endpoint energy of the spectrum, IC denotes internal conversion electrons, neutrons accompany spontaneous fission. d Includes the contribution from annihilation photons and, in the case of spontaneous fission, the prompt and delayed photons e In the case of spontaneous fission, includes the contribution of prompt beta decay.

a

26 m

*235mU

γ 0.0136, α 5.16

α

2.4 × 104 y

Pu

239

γ 0.0173, α 5.46, α 5.499

α SF

87.7 y

Pu

238

β 1.53

β–

4.2 m

*206Tl

γ 0.305, β 1.31

β–

8.15 m

*206Hg

α 5.30

α

138 d

*210Po

β 1.16

β– α

5.01 d

*210Bi

A. RADIOLOGICAL PROPERTIES OF RADIONUCLIDES

/ 201

Appendix B How to Perform a Radiation Survey for Contamination: Instructions for 18 Workers18 In performing a contamination survey with a hand-held instrument, first ensure the instrument is functioning properly. It is advisable to wrap the meter probe with plastic wrap to protect the probe from contamination (except if you are surveying for alpha contamination). Make sure that the instrument has batteries and that they work. To do this, turn your instrument to battery check. If the batteries are acceptable, turn the dial to a measurement mode and use a check source to verify the instrument is operating properly. B.1 Screening Survey If a large population must be surveyed, it is acceptable to perform only a screening survey of the head, face, shoulders and feet rather than a more detailed survey, since these are the most likely locations to become contaminated. You may also consider using portal monitors. If only performing a screening survey for beta-gamma activity, it is acceptable to hold the survey-meter probe ~3 to 5 cm (~1 to 18Adapted

from the Handbook for Responding to a Radiological Dispersal Device. First Responders Guide—The First 12 Hours (CRCPD, 2006).

202

B.2 COMPLETE WHOLE-BODY SURVEY

/ 203

2 inches) away from the body [instead of ~1 cm (0.5 inch) for alpha activity], and move it twice as fast as the normal ~3 to 5 cm s–1 (1 to 2 inches s–1). (If the probe is moved too quickly, its detection capability may be reduced.) Check with local and state radiation-control personnel to determine the extent of contamination survey required. Return the probe to its holder on the meter when finished. Do not set the probe down on the ground. The probe should be placed in the holder with the sensitive side of the probe facing to the side or facing down so that the next person to use the meter can monitor his/her hands without handling the probe or allowing contamination to fall onto the probe surface. B.2 Complete Whole-Body Survey If feasible, perform a complete, whole-body contamination survey and record the findings on the contamination survey sheet. To begin a body survey, the individual should stand with their legs spread and arms extended. First holding the probe about a ~1 cm (0.5 inch) away from the surface to be surveyed, slowly [~3 to 5 cm s–1 (1 to 2 inches s–1)] move the probe over the head, and proceed to survey the shoulders, arms, and bottoms of the feet. Care must be taken not to permit the detector probe to touch any potentially-contaminated surfaces. It is not necessary to perform the personnel contamination survey in exactly the order listed below, but a consistent procedure should be followed to help prevent accidentally skipping an area of the body. Pause the probe for ~5 s at locations most likely to be contaminated: • top and sides of head, face (pause at mouth and nose for ~5 s; high readings may indicate internal contamination); • front of the neck and shoulders; • down one arm (pausing at elbow), turn arm over; • backside of hands, turn over (pause at palms for ~5 s); • up the other arm (pausing at elbow), turn arm over; • shoe tops and inside ankle area; and • shoe bottoms (pause at sole and heel). As with the screening survey, return the probe to its holder on the meter when finished. Do not set the probe down on the ground. The probe should be placed in the holder with the sensitive side of the probe facing to the side or facing down so that the next person to use the meter can monitor his/her hands without handling the probe or allowing contamination to fall onto the probe surface.

204 / APPENDIX B B.3 Most Common Mistakes Made During the Survey • Holding the probe too far away from the surface (should be ~3 to 5 cm (~1 to 2 inches) away for a screening survey or about 0.5 inch or less for an alpha survey). • Moving the probe too fast [should be ~5 to 10 cm s–1 (~2 to 4 inches s–1) for a beta-gamma survey or ~3 to 5 cm s–1 (~1 to 2 inches s–1) for an alpha survey]. • Contaminating the probe; probe background should be observed and compared to initial background. If within a factor of two, it is acceptable to continue to use probe. Otherwise, check with radiation-control personnel. Wrapping the probe in plastic wrap will help prevent surface contamination but this wrapping should not be used when surveying for alpha contamination.

Appendix C How to Distinguish Between Alpha, Beta and Gamma Radiation Using a Geiger-Muller 19 Survey Meter19 This appendix describes a technique using a pancake GeigerMuller (GM) survey meter that may be employed by emergency responders to make a quick, initial determination of the type of radiation (alpha, beta or gamma) present at the scene. Many studies show that the most likely radionuclide(s) to be used in a radiological or nuclear incident would be either a gamma-ray emitter or a beta-gamma emitter. However, it is possible that the radionuclide may be a pure beta emitter such as 90Sr or an alpha emitter such as 239 Pu. This methodology was developed to assist emergency responders in making an initial determination of the type of radiation present. This determination should be used to make decisions until hazardous materials or radiation-control staff arrives at the site with more sophisticated instrumentation to verify the type of radiation and identify the radionuclide(s). Pancake GM survey meters will respond to beta, gamma, and x radiation. They have very limited response to alpha radiation. NaI(Tl) survey instruments will respond primarily to gamma radiation or x rays. Do not be misled into thinking that radionuclides are not present by the lack of response or low reading from a NaI(Tl) survey meter, since it cannot detect alpha and may respond poorly to beta radiation. 19Adapted

from the Handbook for Responding to a Radiological Dispersal Device. First Responders Guide—The First 12 Hours (CRCPD, 2006).

205

206 / APPENDIX C C.1 Determining the Presence of an Alpha-Emitting Radionuclide Using Only a Geiger-Muller Survey Meter Although alpha emitters may not appear to be as hazardous as gamma-ray emitters, they are very harmful when inhaled or ingested. Therefore, it is important to check for the presence of alpha-emitting radionuclides. Because the instruments normally available to emergency responders will not readily respond to alpha particles, it is important to use appropriate respiratory protection when monitoring for radionuclides. Procedure: • Take readings at approximately 8 cm (3 inches) and 1 cm (0.5 inch) (as close as possible without touching) above the potentially-contaminated surface with the window facing down. • If the instrument reading increases by more than a factor of three at the 1 cm (0.5 inch) measurement [as compared to the 8 cm (3 inch) measurement], suspect alpha contamination (such as 239Pu). • Next, place a sheet of paper on the surface and take a reading with the window side down directly on top of the paper. The alpha radiation will not penetrate the paper, and the window-down reading should significantly decrease to near background level. If the window-down measurement taken over the paper does not significantly decrease, the radionuclide is likely not an alpha emitter. (Note that some alpha emitters, such as 241Am, also emit a low-energy gamma which will not be stopped by a sheet of paper.) C.2 Determining the Presence of Strontium-90 (or other pure beta emitters) Using a Pancake Geiger-Muller Survey Meter Strontium-90 is a pure beta emitter, and most NaI(Tl) instruments respond poorly to energetic beta emitters. However, 90Sr beta radiation can be easily detected and measured with a survey meter connected to an end-window, side-window, or pancake GM probe (preferred). For the purposes of this Report, suspect the presence of 90 Sr if a pancake GM survey meter reads high, such as between 1,000 and 10,000 cpm (20 to 200 times background) but there is no corresponding increase in readings using a NaI(Tl) survey meter (still reads near background). When 90Sr is shielded by certain materials, the beta radiation cannot be detected. However, the interaction of the beta radiation

C.2 DETERMINING THE PRESENCE OF STRONTIUM-90

/ 207

with the shielding materials can produce x rays, which can be detected by GM, NaI(Tl), and other types of gamma-ray identification survey meters. Procedure: • Take a measurement with the window side of the pancake probe (mesh covered side) facing down at ~15 cm (~6 inches) above the area being surveyed where the survey meter reads between 500 to 1,500 cpm. Then take another measurement with the window side facing up (away from the ground) at the same height. • Compare the two measurements: - If only 90Sr (or another pure beta-emitter or a very weak gamma emitter such as 241Am) is present, the window-up reading will be near background (depending on the model of the GM pancake probe, background should be in the range of 25 to 75 cpm), and the window-down reading should be 10 or more times greater than the window-up reading. This is because the beta emissions are not able to penetrate the back side of the GM pancake probe. - If an energetic gamma emitter is present (e.g., 137Cs, 192Ir, 60Co), the window-down reading at 15 cm (6 inches) will be approximately twice the window-up reading. • Take another measurement with the window side of the pancake probe facing down at ~1 m (~3 feet) from the area being surveyed where the meter reads between 500 to 1,500 cpm. Then take another measurement with the window-up at the same height. Compare the two measurements. If a gammaemitting radionuclide is present, both readings will be approximately the same.

208

20

Middle initial:

Drivers license #:

Phone:

Last name:

Reproduced with minor changes from the Handbook for Responding to a Radiological Dispersal Device (CRCPD, 2006).

Parent or guardian (if child):

Location at time of incident:

Date/time:

Address:

Date of birth:

First name:

D.1 Contamination Survey Sheet20

Survey and Registry Forms

Appendix D

Survey instrument and detector type:

<1,000 cpm

Survey results

Front

>1,000 cpm

Back

>10,000 cpm

Mark contamination locations and survey reading on the diagrams below (circle if readings are in cpm mR h–1 μR h–1)

D.1 CONTAMINATION SURVEY SHEET

/ 209

Person sent to medical facility:

Yes

Yes

Nasal area reading of 100,000 cpm (0.5 mR h–1): No

Valuables returned:

Yes

Clothing and valuable bag number:

Person sent to decontamination area:

Monitored by:

Comments:

Yes

No

No

(if yes, refer to medical facility)

No

210 / APPENDIX D

Last

Yes

Possibly pregnant:

No

No

Sex:

Other (specify)

If yes, estimate term:

Place of birth:

First

Emergency services Yes

Public

Witness to the incident:

Member of:

Email address:

Telephone:

Current permanent full address

Social security #:

Nationality:

Date of birth:

Full name:

Date:

D.2 Registry Form

M

M.I. F

D.2 REGISTRY FORM

/ 211

<1 μSv h–1

Background reading

Yes

Yes

No

No

Scheduled for follow-up:

Yes

No

Medical triage category: (based on the medical condition)

Field decontamination:

Decontamination procedures performed:

Model

Radiological survey performed:

Time spent at each location:

Location(s) during emergency:

need immediate treatment need early treatment can wait for treatment no need for treatment

Priority 2: Priority 3: Priority 4:

Yes

Priority 1:

Full decontamination:

>1 μSv h–1

Personal survey instrument

Instrument type:

No

212 / APPENDIX D

Remarks:

Organization:

Telephone:

Date:

Full name: Time:

D.2 REGISTRY FORM

/ 213

214 / APPENDIX D D.3 ATSDR Rapid Response Registry Survey Form

D.3 ATSDR RAPID RESPONSE REGISTRY SURVEY FORM

/ 215

Appendix E How to Perform Decontamination at 21 Home21 You may have been exposed to low levels of radioactive particles. The particles may have settled on your hair, skin and clothing as dust. You are not in immediate danger from these small radioactive particles. However, you should go home or to a facility designated by the emergency operations center to decontaminate. Removal of outer clothing should reduce your contamination by up to 90 %. In order to help protect your health and safety as well as others, please follow these directions. Because radiation cannot be seen, smelled, felt or tasted, people at the site of an incident will not immediately know if you have been exposed to radioactive materials. You can take the following steps to limit your contamination: • Get out of the immediate area quickly. Go directly home, inside the nearest safe building, or to an area to which you are directed by law-enforcement or health officials. Do not go to a hospital unless you have a medical condition that requires treatment. • If radioactive material is on your clothes, removing them will reduce the external contamination and decrease the risk of internal contamination. Prompt removal of outer clothing will also reduce the length of time that you are exposed to radiation. When removing the clothing, be careful of any clothing that has to be pulled over the head. Try to 21Adapted

from the Handbook for Responding to a Radiological Dispersal Device. First Responders Guide—The First 12 Hours (CRCPD, 2006).

216

E. HOW TO PERFORM DECONTAMINATION AT HOME

•

•

•

•

/ 217

either cut the article off or prevent the outer layer from coming in contact with the nose and mouth area. You may also hold your breath while carefully pulling the article over the head. Removal of clothes should be done in a garage or outside area if available, where the ground can be washed with a hose. If an outside area is not available, the removal of clothing should take place in a room where the floor can be easily cleaned, such as the tub or shower areas. (Disposable sweeping cloths are good for decontaminating smooth floor surfaces.) Clothing should be rolled up with the contaminated side “in” to minimize cross contamination. If possible, place the clothing in a plastic bag (double-bagging is best to reduce the chances of a rupture), and leave it in an out-of-the-way area, such as the corner of a room or garage. Keep people away from it to reduce their exposure to radiation. You may be asked to bring this bag for follow-up readings or for disposal at a later time. Keep cuts and abrasions covered when handling contaminated items to avoid getting radioactive material in the wound. Shower and wash all of the exposed parts of your body and hair using lots of soap and lukewarm water to remove contamination. This process is called decontamination. Simple washing will remove most of the radioactive particles. Do not use abrasive cleaners, or scrub too hard. Do not use hair conditioners. If you are going to a monitoring location, it is best to change clothes and shower before being monitored. Contact your local or state radiation-control program for additional guidance.

Appendix F Using Geiger-Muller Survey Meters to Assess Internal Contamination for Selected GammaEmitting Radionuclides Section 8.5 describes a method for rapid screening of patients to determine whether internal contamination exceeds the CDG. Tables F.1, F.2, and F.3 are based on those developed by Hurtado (2006) and Juneja (2011), and the data currently are available from CDC (2009). The values in these tables are based on the response of a typical thinwindow (“pancake”) detector [Ludlum 44-9® (Atlanta Nuclear, Rockland, Massachusetts) pancake GM probe]. For each particular radionuclide, inhalation exposures are based on a 1 μm AMAD particle size and Type-M lung absorption rate in the case of 60Co and 192Ir and Type-F lung clearance rate in the case of 137Cs. The distance for checking the patient was fixed at 6 and 30 cm directly in front of the sternum (AP) or directly in the middle of the back (PA). The numerical values indicated on each table below are the net counts per minute that would indicate an intake at the CDG level, defined as the intake of radionuclide that would result in an effective dose of 250 mSv for an adult or 50 mSv for a child. Results given for a child are based on calculations for a 10 y old child (Section 8.5). If the indicated count rate exceeds the tabulated value, it should be assumed that the CDG has been met or exceeded. Consult CDC (2009) to determine whether tables have been updated or whether tables for additional radionuclides, ages or distances have been added. This screening method assumes that the patient is free of external contamination. If a reading from the chest (AP) indicates that a patient has an intake in excess of the CDG, it would be prudent to measure from the back (PA) position to confirm that the indicated activity is from an intake rather than external contamination. 218

F. USING GEIGER-MULLER SURVEY INSTRUMENTS

/ 219

TABLE F.1a—Adult male count rate (counts per minute) corresponding to 1 CDG for inhaled 60Co (Type M, 1 μm AMAD) using a GM survey meter (CDC, 2009). Distance from Sternum or Middle of Spine Time Post Incident (h)

6 cm AP

30 cm PA

AP

PA

1

12,900

9,200

4,800

3,100

2

11,300

8,500

4,600

2,900

4

10,700

8,100

4,500

2,800

6

10,600

8,000

4,400

2,800

8

10,600

7,800

4,300

2,700

10

10,500

7,700

4,200

2,700

12

10,400

7,600

4,100

2,600

14

10,200

7,500

4,000

2,500

16

10,000

7,400

3,900

2,500

18

9,800

7,300

3,800

2,400

20

9,600

7,200

3,700

2,400

24

9,100

7,000

3,400

2,300

48

6700

6,100

2,200

1,800

72

5,500

5,600

1,700

1,500

220 / APPENDIX F TABLE F.1b—Adult female count rate (counts per minute) corresponding to 1 CDG for inhaled 60Co (Type M, 1 μm AMAD) using a GM survey meter (CDC, 2009). Distance from Sternum or Middle of Spine Time Post Incident (h)

6 cm AP

30 cm PA

AP

PA

1

15,500

9,600

5,400

3,200

2

14,500

9,100

5,100

3,000

4

13,500

8,600

4,900

2,900

6

13,000

8,300

4,800

2,800

8

12,800

8,100

4,700

2,700

10

12,500

7,900

4,500

2,600

12

12,000

7,700

4,400

2,500

14

11,800

7,600

4,200

2,400

16

11,500

7,400

4,100

2,400

18

11,200

7,300

4,000

2,300

20

11,100

7,100

3,800

2,300

24

10,500

6,900

3,500

2,100

48

8,200

5,900

2,300

1,700

72

7,300

5,400

1,800

1,400

F. USING GEIGER-MULLER SURVEY INSTRUMENTS

/ 221

TABLE F.1c—Child count rate (counts per minute) corresponding to 1 CDG for inhaled 60Co (Type M, 1 μm AMAD) using a GM survey meter (calculated from data in CDC, 2009). Distance from Sternum or Middle of Spine Time Post Incident (h)

6 cm AP

30 cm PA

AP

PA

1

2,500

870

1,600

520

2

2,300

820

1,500

490

4

2,200

790

1,400

460

6

2,100

770

1,300

440

8

2,100

750

1,300

430

10

2,000

730

1,300

420

12

2,000

700

1,200

400

14

1,900

680

1,200

390

16

1,800

660

1,200

380

18

1,800

640

1,200

370

20

1,700

610

1,100

360

24

1,600

570

1,100

340

48

1,300

380

950

270

72

1,200

290

880

230

222 / APPENDIX F TABLE F.2a—Adult male count rate (counts per minute) corresponding to 1 CDG for inhaled 137Cs (Type F, 1 μm AMAD) using a GM survey meter (CDC, 2009). Distance from Sternum or Middle of Spine Time Post Incident (h)

6 cm AP

30 cm PA

AP

PA

1

6,300

3,800

2,300

1,500

2

5,700

3,700

2,200

1,400

4

5,000

3,500

2,000

1,400

6

4,700

3,400

1,900

1,300

8

4,400

3,300

1,800

1,300

10

4,100

3,200

1,700

1,300

12

3,900

3,100

1,600

1,200

14

3,700

3,000

1,600

1,200

16

3,500

3,000

1,500

1,200

18

3,400

2,900

1,500

1,200

20

3,300

2,900

1,400

1,100

24

3,100

2,800

1,300

1,100

48

2,700

2,600

1,100

1,000

72

2,500

2,500

1,000

1,000

F. USING GEIGER-MULLER SURVEY INSTRUMENTS

/ 223

TABLE F.2b—Adult female count rate (counts per minute) corresponding to 1 CDG for inhaled 137Cs (Type F, 1 μm AMAD) using a GM survey meter (CDC, 2009). Distance from Sternum or Middle of Spine Time Post Incident (h)

6 cm AP

30 cm PA

AP

PA

1

10,000

5,000

3,200

1,700

2

9,600

4,900

3,000

1,700

4

8,300

4,400

2,800

1,600

6

7,200

3,900

2,600

1,500

8

6,300

3,400

2,400

1,400

10

5,500

3,100

2,200

1,300

12

4,900

2,800

2,100

1,300

14

4,300

2,600

2,000

1,200

16

3,900

2,400

1,900

1,200

18

3,600

2,300

1,800

1,200

20

3,300

2,200

1,700

1,100

24

2,800

2,000

1,600

1,100

48

1,800

1,600

1,300

1,000

72

1,600

1,500

1,200

900

224 / APPENDIX F TABLE F.2c—Child count rate (counts per minute) corresponding to 1 CDG for inhaled 137Cs (Type F, 1 μm AMAD) using a GM survey meter (calculated from data in CDC, 2009). Distance from Sternum or Middle of Spine Time Post Incident (h)

6 cm AP

30 cm PA

AP

PA

1

2,300

730

1,200

400

2

2,200

700

1,100

390

4

1,900

640

1,000

360

6

1,700

590

900

340

8

1,500

550

800

330

10

1,300

520

720

310

12

1,100

490

660

300

14

1,000

460

610

290

16

910

440

570

280

18

830

420

530

280

20

770

410

510

270

24

660

380

460

260

48

430

300

380

230

72

370

270

360

220

F. USING GEIGER-MULLER SURVEY INSTRUMENTS

/ 225

TABLE F.3a—Adult male count rate (counts per minute) corresponding to 1 CDG for inhaled 192Ir (Type M, 1 μm AMAD) using a GM survey meter (CDC, 2009). Distance from Sternum or Middle of Spine Time Post Incident (h)

6 cm AP

30 cm PA

AP

PA

1

14,400

9,500

6,000

3,300

2

12,500

8,700

5,600

3,000

4

12,000

8,200

5,500

2,900

6

12,000

8,000

5,400

2,800

8

12,000

7,900

5,300

2,800

10

12,000

7,800

5,200

2,800

12

11,600

7,700

5,100

2,700

14

11,500

7,600

5,000

2,700

16

11,300

7,500

4,900

2,600

18

11,100

7,500

4,700

2,600

20

10,800

7,400

4,500

2,500

24

10,300

7,200

4,200

2,400

48

7,600

6,500

2,700

2,000

72

6,100

6,000

2,000

1,700

226 / APPENDIX F TABLE F.3b—Adult female count rate (counts per minute) corresponding to 1 CDG for inhaled 192Ir (Type M, 1 μm AMAD) using a GM survey meter (CDC, 2009). Distance from Sternum or Middle of Spine Time Post Incident (h)

6 cm

30 cm

AP

PA

AP

PA

1

17,800

10,200

6,600

3,400

2

16,300

9,500

6,200

3,200

4

15,000

9,000

5,900

3,000

6

14,500

8,500

5,700

2,800

8

14,000

8,200

5,600

2,700

10

13,500

8,000

5,400

2,600

12

13,200

7,800

5,300

2,600

14

12,800

7,700

5,100

2,500

16

12,500

7,500

5,000

2,400

18

12,200

7,400

4,800

2,400

20

11,900

7,300

4,600

2,300

24

11,300

7,200

4,300

2,200

48

9,200

6,400

2,800

1,800

72

8,200

6,000

2,100

1,600

F. USING GEIGER-MULLER SURVEY INSTRUMENTS

/ 227

TABLE F.3c—Child count rate (counts per minute) corresponding to 1 CDG for inhaled 192Ir (Type M, 1 μm AMAD) using a GM survey meter (calculated from data in CDC, 2009). Distance from Sternum or Middle of Spine Time Post Incident (h)

6 cm AP

30 cm PA

AP

PA

1

2,900

1,100

1,700

560

2

2,700

1,000

1,500

530

4

2,500

960

1,400

490

6

2,400

940

1,400

460

8

2,300

910

1,300

440

10

2,200

890

1,300

430

12

2,100

470

1,300

420

14

2,100

840

1,200

410

16

2,000

810

1,200

400

18

2,000

780

1,200

390

20

1,900

760

1,200

380

24

1,800

700

1,100

370

48

1,500

460

1,000

300

72

1,300

350

980

260

Appendix G Collection and Preparation of Biological Samples for Radioanalysis G.1 Urine Samples Urine analysis is particularly useful for those radionuclides which do not emit types of radiation that are sufficiently penetrating to be detected outside the body. Analysis of the amount of the radionuclide in urine can be done by measuring activity (such as for cesium) or by chemical analysis (such as for uranium). By use of a biokinetic model, the radionuclide concentration in urine and time of intake are used to estimate the body burden of the radionuclide and its distribution among different organs that may follow internal deposition via different routes of exposure. Figure 3.1 is a generic biokinetic diagram for internally-deposited radionuclides including the various routes by which these radionuclides can enter the body and eventually be excreted. G.2 Main Collection Issues • Discard first void: Void is the term used to mean a single relief of the bladder such that a person empties all the available urine from their body. Generally, the first void after a potential intake should be discarded. It should not be collected as part of the sample to be measured for two primary reasons. First, typically there has not been enough elapsed time for the radioactive material to pass from where it enters the body, into the blood, into the kidney, and into the bladder. Second, if it does enter a partially filled bladder then the radioactive material will be diluted and the sample may result in the underestimation of the intake. 228

G.2 MAIN COLLECTION ISSUES

/ 229

• Minimum time to wait before collecting sample: There exists some finite amount of time between inhalation, ingestion, absorption or injection of a radionuclide and it appearing in urine. The time it takes is variable and depends on many factors including method of entry, biokinetics of the individual, radionuclide involved, chemical form of the radionuclide, and physical form of the radionuclide. Generally, waiting 4 h from the time of probable entry to the start of collecting a 24 h sample is sufficient to obtain the measurement. • Period of collection: Ideally, urine should be collected over a 24 h period. This kind of sample is called a “24 h urine sample.” Collecting a sample for a shorter time or from a single spot sample may result in a less accurate determination of intake activity. NCRP Report No. 87 (NCRP, 1987) states the following: “Although various kinetic models relate urinary excretion to the amount of a radionuclide in the body, the actual data on which most of these models are based exhibit considerable variability among samples from a single individual or from different individuals (Pochin, 1968; Snyder et al., 1972). Thus, a single spot sample taken from a person usually has a high degree of physiologically related uncertainty and does not provide a reliable estimate of body burden.” With that stated, sometimes it is only possible to collect a spot sample due to circumstances. In these cases, the first void should especially be discarded, if possible. When it is only possible to get a spot sample of urine, a correction should be made to scale the sample result up to an approximate 24 h sample for evaluation. More detailed discussion on this scaling procedure is provided in Section 7.2.3 of this Report and Section 10.3.1.2 of NCRP Report No. 161 (NCRP, 2008a). • The volume of urine collected from an individual over a 24 h period will vary somewhat depending on the person and situation. Therefore, the measurement of activity in the sample is normalized to remove the volume dependency. The 24 h reference urine volumes given in ICRP Publication 89 are 1.6 L for men and 1.2 L for women (ICRP, 2002b). Sometimes a 24 h urine sample is incorrectly interpreted to mean a sample collected 24 h after the potential intake. This distinction should be made clear when communicating with people who are unfamiliar with the concept of a 24 h

230 / APPENDIX G urine sample. • Cross-contamination: The concern for cross-contamination is that radioactive material external to the body may end up in the urine sample during collection of the urine. When a low level of contaminate is to be measured, cross-contamination can significantly affect the accuracy of the measurement (Sun et al., 1993). To prevent cross-contamination it is recommended that at the beginning of the 24 h collection period and before collecting the void, the person giving the sample should shower and put on clean clothes. Also, the container used to hold the urine should be kept closed until needed. • Containers: Containers for collecting a 24 h urine sample should hold 2 L or more since reference 24 h urinary excretion is 1.6 L for men and 1 L for women. Commercially obtained containers are readily available for this purpose and have openings that are wide enough to make collection easy for both men and women. In a catastrophic incident an empty plastic 2 L soda bottle could be used. If possible, it should be rinsed and dried. Because the opening is too small to be practically used it may be necessary to urinate into a wider container, such as a measuring cup, and then transfer the urine to the larger 2 L soda bottle. • Documentation: Tracking the urine sample with the proper documentation is critical to maintaining the chain of custody and to the proper determination of the body burden. • Personal information: Include, for example, name, age, sex, address, telephone number, email address, and social security number (or at least the last four digits of the social security number if possible). • Technical information: Include, for example, date/time of probable intake, start date, time and duration of the urine collection, number of voids, if first void was discarded. • Other considerations: Clean containers must be used for collection and storage. • All biological samples: Subject to deterioration by bacteriologic action that may interfere with subsequent analysis. Prompt analysis following collection is preferred. When samples are kept longer then a day, they should be refrigerated, acidified to minimize precipitation, or have preservatives added to inhibit bacterial growth. Appropriate handling techniques should be used to avoid exposure to possible diseases like hepatitis.

Appendix H Shipping of Biological Samples H.1 Introduction It may be necessary to send a biological sample for testing and analysis. Possible types of biological samples could include nasal swabs, sputum, urine, or feces. Hospital or clinical laboratory personnel most likely are aware of shipping regulations and have materials on hand to ship specimens to the analytical laboratory that will be performing the analysis. Additionally, the analytical laboratory should be consulted for shipping instructions. The following details are provided for background information and for those instances when neither the hospital or clinical laboratory nor the analytical laboratory is able to provide instructions. However, readers should be aware that the shipping requirements provided here may have changed subsequent to publication of this Report. The commercial carrier should be contacted any time there are questions about shipping requirements. There is a distinction between shipping clinical biological samples and hazardous biological samples. This appendix is written to assist in the preparation and shipping of clinical samples for routine testing and not those biological samples that may contain infectious diseases. Sometimes clinical biological samples are referred to as diagnostic specimens or exempt biological samples. Sources for detailed information on shipping biological samples include the U.S. Department of Transportation (DOT, 2009), Federal Express® (Memphis, Tennessee) (FedEx, 2010), Glode and Gillum (2010), the International Air Transport Authority (IATA, 2010), and USPS (2009). H.2 Regulatory Information: Brief Summary DOT regulates transport within U.S. borders. Title 49 of the Code of Federal Regulations (49 CFR) has specific requirements 231

232 / APPENDIX H for packaging and labeling hazardous goods. In addition, DOT requires that those who ship or receive hazardous goods receive both awareness training (about the reasons for regulation) and job function (about how to properly prepare a shipment). The United Nations has a committee of experts (International Civil Aviation Organization), which has harmonized international shipping regulations; United Nations’ “certified” packaging has been tested and certified to meet criteria for safe transport of hazardous goods. The International Air Transport Authority (IATA), an aviation trade organization, absorbs the International Civil Aviation Organization technical instructions into IATA regulations. Noncompliance with regulations may result in large fines, and in some instances (e.g., concealing cultures in personal baggage on an airline), personal civil and criminal liability may result. H.3 Definitions These definitions are according to DOT regulations 49 CFR Part 173.134 (DOT, 2009). Similar definitions are in IATA (2010). • Diagnostic/exempt specimen (e.g., urine samples being shipped for routine testing not related to diagnosis of an infectious disease are exempt from shipping requirements for infectious disease specimens): Any human or animal material being shipped not known or suspected of containing a pathogen. This includes secreta, excreta, blood, blood components, tissue, and body fluids. For example, blood samples sent for a cancer screen are considered diagnostic specimens but blood samples sent to confirm HIV seroconversion are classified as infectious substances because the presence of a pathogen is suspected. Diagnostic specimens must be shipped with the triple packaging explained below. • Biological product: Material prepared and manufactured in accordance with certain regulations of the U.S. Department of Agriculture or DHHS (e.g., vaccine preparations). • Infectious substance: In hazardous goods regulations, “infectious substance” includes any toxin (classified as 6.1) or infectious agents (classified as 6.2) which affect humans or animals. This includes infectious agents listed in 42 CFR Part 72.3 of the regulations of DHHS and any other agent that causes or may cause severe, disabling or fatal disease. H.4 General Shipping Guidelines for Clinical, Diagnostic and Exempt Biological Samples • All shipments must include four basic requirements:

H.4 GENERAL SHIPPING GUIDELINES

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- watertight primary containers - watertight secondary containers - absorbent material - sturdy outer packaging • Examples of suitable watertight primary containers include the following: - plastic canister - glass/plastic vial - sealed plastic bag - glass/plastic jar Note: Use sealed plastic bags for tissue and solid samples only. • Examples of suitable watertight secondary containers include the following: - screw-cap can - sealed Styrofoam® (Dow Chemical, Midland, Michigan) container [minimum of 2.5 cm (1 inch) thick] - sealed plastic bag - plastic container • Examples of suitable absorbent materials include the following: - super absorbent packer - cellulose wadding - cotton balls - paper towels Note: Absorbent materials must be placed between the primary and secondary container. The quantity should be sufficient to absorb all liquid in the shipment. • Examples of sturdy outer packing include the following: - corrugated fiberboard - wood - rigid cooler - rigid plastic container Note: The packaging must be of good quality (typically capable of passing a drop test of 1.2 m) and must be marked on the outside with a label indicating it is a clinical biological sample (such as the words exempt human specimen or diagnostic specimens). Corrugated cardboard is the usual choice. Styrofoam® boxes, plastic bags, or paper envelopes are unacceptable outer containers for shipping biological materials. Exact packing requirements and labeling are specific to the shipper chosen.

234 / APPENDIX H H.5 Specific Shipping Guidelines for Clinical, Diagnostic and Exempt Biological Samples Due to the highly specific nature of shipping regulations, specific details for shipping patient specimens are not provided here. The clinical or hospital laboratory or emergency department where biological specimens are collected should follow their standard protocol when shipping these specimens to the analytical laboratory that will be performing the radioactivity analysis. It is highly recommended that the analytical laboratory and commercial carrier be contacted for any special packaging and shipping instructions. H.6 Website Links for Some Commercial Shippers • DHL® (Germany): http://www.dhl-usa.com; 1-800-225-5345 • Federal Express® (Memphis, Tennessee): http://www.fedex. com; 1-800-Go-FedEx. Additional assistance for noninfectious shipment inquiries is available from FedEx® Packaging Design and Development Department at 1-800-633-7019 • United Parcel Service® (Atlanta, Georgia): http://www.ups. com; 1-800-554-9964 • World Courier® (Stamford, Connecticut): http://www.worldcourier.com; 1-800-221-6600

Appendix I Population Screening and Monitoring Implications of Two Urban Contamination Incidents

I.1 Background Two contamination incidents provide some objective evidence on the level of response that may follow release of radioactive contamination in an urban environment. The two incidents are the 1987 widespread release of 137Cs in urban Goiânia, Brazil, and the November 2006 210Po incident in London. In 1987, accidental dispersal of 137Cs resulted from the actions of unauthorized salvagers breaking open an abandoned teletherapy machine in Goiânia. In the London case, the focus was a very limited intentional administration to a single individual, but it was not discovered for some time after the poisoning and contamination spread to a number of individuals that required the emergency response to be scaled up. In Goiânia, the initial contamination was not recognized or suspected by the involved individuals and ultimately became widespread in an urban area. The follow-up actions to both incidents might be comparable to those likely to follow the use of an RDD or IND. Comparison of the scenarios includes the nature of the radionuclides and exposure scenario, the extent of contamination spread, the resulting population monitoring and bioassays, the number of confirmed intakes and potential doses, and the ameliorative actions undertaken.

235

236 / APPENDIX I I.2 Overview of Two Cases I.2.1

Goiânia, Brazil

The Goiânia incident involved the unrecognized breaching by unauthorized salvagers of a 50.9 TBq (1,375 Ci) 137Cs source from an abandoned teletherapy machine (IAEA, 1988). Cesium-137 emits energetic beta particles and relatively-high-energy gamma rays, making the radionuclide quite easy to detect using portable survey instruments (GM detectors) and gamma scintillator [NaI(Tl)] detector systems. In addition to the hazard posed by the highly penetrating gamma-ray radiation and the energetic beta particles, the soluble cesium chloride salt form was readily dispersible and subject to intake by ingestion and inhalation. Once intake occurred, the biokinetics of 137Cs resulted in a rapid distribution throughout the body causing essentially a total body distribution and irradiation from internally-deposited radioactive material. Within 2 d, one of the salvagers sought medical attention for what was thought initially to be an allergic reaction to bad food, but was in actuality a manifestation of the ARS. It was several additional patients and almost two weeks before radiation exposure was recognized as a likely agent. During this time, contamination was spread over a large urban area. Ultimately, four patients died from their exposure and an additional 249 others showed external contamination with 137Cs. A total of 46 persons underwent medical therapy using Prussian blue for decorporation of 137Cs. Approximately 112,000 members of the public were screened for contamination by personal survey. I.2.2

London, United Kingdom

The radionuclide poisoning case in London provides an example of how the apparent internal contamination of a single person may require subsequent scaling of response to monitor a number of people who may have come in contact with the radionuclide. This case demonstrates response to a radiological incident that is discovered days after the initial exposure. This case also demonstrates the impact on effective screening of a population that is facilitated by good communication, effective and timely press releases, and excellent efforts to monitor and report back to members of the public. Polonium-210 was used as the agent in an assassination on November 1, 2006 in London, United Kingdom (Harrison et al., 2007). As of this writing, the crime is still under investigation and many of the details are not publicly available. The nature of 210Po is unusual in that it is the only alpha-emitting radionuclide that is distributed broadly when taken into the body in a relatively soluble form. This

I.2 OVERVIEW OF TWO CASES

/ 237

distribution pattern included alpha radiation exposure of the bone marrow that resulted in the symptoms of the ARS. It was just a matter of days before the victim died (November 23, 2006) that radiation poisoning with 210Po was diagnosed. The diagnosis was made following recognition that a large-volume urine sample appeared to be emitting radiation. This was detected by a passing hospital radiation technologist using a GM survey meter. The low-yield gamma rays emitted by 210Po were of sufficient quantity to elevate the background reading. Once the radionuclide was suspected, surveys quickly revealed its presence, not only in the patient, but also on objects with which he had contact. Although only one person was the target of the assassination, contamination was found in various locations, including offices, restaurants, coffee bars, night clubs, soccer stadium, airplanes, cars, and three hotels. Literally, thousands of people were potentially exposed, including 460 overseas visitors from 52 countries outside the United Kingdom. Monitoring and bioassay of those inside the United Kingdom consisted of announcing that concerned citizens should contact the National Health Service, which then determined potentials for exposure and need for follow-up based on a triage script from the Health Protection Agency. The Health Protection Agency referred overseas visitors to their respective national health authorities for consideration of follow-up. Where indicated, follow-up consisted of a 24 h urine sample radiochemically analyzed for 210Po. The action levels established were based on committed effective dose levels of <1 mSv indicating no significant exposure, between 1 and 6 mSv indicating exposure had occurred as a result of a criminal act but the dose was of no concern, and ≥6 mSv as indicating a dose of some concern for increased risk of cancer. As of April 2007, 738 people had submitted urine samples, with results indicating that 686 did not receive significant exposures, 35 were exposed but of no health concern, and 17 were in the category of some concern for increased risk. Those 17 were either hotel or restaurant staff or a family member. Of particular interest is that out of 78 healthcare workers associated with the patient, 77 were in the no significant-exposure category, one was in the middle category, and none were in the highest exposure category. This is a good indication that the standard precautions used by healthcare workers were effective at limiting intakes of activity.

Appendix J Pregnancy Categories for Drug Use TABLE J.1—Current categories for drug use in pregnancy (Meadows, 2001). Category

Description

A

Adequate, well-controlled studies in pregnant women have not shown an increased risk of fetal abnormalities.

B

Animal studies have revealed no evidence of harm to the fetus, however, there are no adequate and well-controlled studies in pregnant women. or Animal studies have shown an adverse effect, but adequate and well-controlled studies in pregnant women have failed to demonstrate a risk to the fetus.

C

Animal studies have shown an adverse effect and there are no adequate and well-controlled studies in pregnant women. or No animal studies have been conducted and there are no adequate and well-controlled studies in pregnant women.

D

Studies, adequate well-controlled or observational, in pregnant women have demonstrated a risk to the fetus. However, the benefits of therapy may outweigh the potential risk.

X

Studies, adequate well-controlled or observational, in animals or pregnant women have demonstrated positive evidence of fetal abnormalities. The use of the product is contraindicated in women who are or may become pregnant.

238

Appendix K Emergency Phone Numbers for Government Officials to Request Assistance Following a Radiological or Nuclear Incident

Agency

Phone

State Radiological Emergency Contact (state number to be filled in by local emergency response personnel) CDC Report a Radiological Emergency

1-770-488-7100

DOE Federal Radiological Monitoring and Assessment Center (FRMAC)

1-800-dial-DOE

EPA Radiological Emergency Response Team (RERT)

1-800-424-8802

Radiation Emergency Assistance Center/Training Site (REAC/TS)

1-865-576-1005

239

Glossary absorbed dose (D): Quotient of dε by dm, where dε is the mean energy imparted by ionizing radiation to matter in a volume element and dm is the mass of matter in that volume element: D = dε /dm. For purposes of radiation protection and assessing dose or risk to humans in general terms, the quantity normally calculated is the mean absorbed dose in an organ or tissue (T): DT = ε T ⁄ m T , where ε is the total energy imparted in an organ or tissue of mass mT . The SI unit of absorbed dose is the joule per kilogram (J kg–1), and its special name is the gray (Gy). In conventional units often used by state and federal agencies, absorbed dose is given in rad; 1 rad = 0.01 Gy. accident: An unintentional or unexpected happening that is undesirable or unfortunate, especially one resulting in injury, damage, harm or loss. activity (A): Rate of transformation (or disintegration or decay) of radioactive material. The SI unit of activity is the reciprocal second (s–1) (meaning one transformation per second), and its special name is the becquerel (Bq). In conventional units often used by state and federal agencies, activity is given in curies (Ci); 1 Ci = 3.7 × 1010 Bq. activity median aerodynamic diameter (AMAD): The diameter of a unit density sphere with the same settling velocity in air as that of an aerosol particle whose activity is the median for the entire aerosol. Fifty percent of the activity (aerodynamically classified) in the aerosol is associated with particles greater than the AMAD. A log-normal distribution of particle sizes is usually assumed. Used when deposition depends principally on impaction and sedimentation. acute radiation syndrome (ARS): A broad term used to describe a range of signs and symptoms that reflect severe damage to specific organ systems that can lead to death within hours or several weeks. aerosol: Any system of liquid droplets or solid particles dispersed in air, of fine enough particle size, and consequent low settling velocity, to possess considerable stability as an aerial suspension. air kerma (kerma) (kinetic energy released per unit of mass) (K): The quotient of the sum of the initial kinetic energies of all the charged particles liberated by uncharged particles in matter divided by the mass of the matter into which the particles are released and is given the special name gray (Gy). 1 Gy = 1 J kg–1. In the event that the matter is air, kerma is often referred to as air kerma. algorithm: A formula used to compute numerical values from an equation, based on step-by-step numerical computations, as opposed to an algebraic computation. Numerical algorithms implemented with computer software are particularly useful for solving complex differential equations or sets of equations that may lack algebraic solutions. alpha radiation: Energetic nuclei of helium atoms, consisting of two protons and two neutrons, emitted spontaneously from nuclei in the decay of some radionuclides. Alpha radiation is weakly penetrating, and can be stopped by a sheet of paper or the outer dead layer of skin. Also

240

GLOSSARY

/ 241

called alpha particle and sometimes shortened to alpha (e.g., alphaemitting radionuclide). Alpha particles may represent a hazard when radionuclides are deposited inside the body (e.g., via inhalation, ingestion or wounds). background radiation: Radiation from cosmic sources, naturally-occurring radioactive material in the earth and in the body (including exposure from radon), and global fallout as it exists in the environment from past testing of nuclear explosive devices. The typically quoted annual average effective dose to an individual from all sources of background radiation is ~3 mSv (~300 mrem). This value includes a major contribution from radon, which is ~2 mSv (~200 mrem). becquerel (Bq): The special name for the unit of activity in the SI system [i.e., one nuclear transformation per second (s–1)]. The conventional unit of activity is the curie (Ci); 3.7 × 1010 Bq = 1 Ci. beta radiation: Energetic electrons or positrons (positively charged electrons) emitted spontaneously from nuclei in decay of some radionuclides. Also called beta particle and sometimes shortened to beta (e.g., beta-emitting radionuclide). bioassay: Any procedure used to determine the nature, 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. Generally used for the purpose of estimating intake and committed dose. biodosimetry: Use of clinical and laboratory observations to estimate radiation dose received after radiation exposure. biokinetics: The time course of absorption, distribution, metabolism and excretion of a substance introduced into the body of an organism. bremsstrahlung: Secondary photon radiation produced by deceleration of charged particles passing through matter. calibration: For an instrument intended to measure dose or dose-raterelated quantities, calibration is the determination of the instrument response in a specified radiation field delivering a known dose (rate) at the instrument location; calibration normally involves the adjustment of instrument controls to read the desired dose (rate) and typically requires response determination on all instrument ranges. catastrophic incident: “Any natural or man-made incident, including terrorism, that results in extraordinary levels of mass casualties, damage, or disruption severely affecting the population, infrastructure, environment, economy, national morale, and/or government functions” (FEMA, 2009). chelate: Chemical compound in which the central atom (usually a metal ion) is attached to neighboring atoms by at least two bonds in such a way as to form a ring structure. chelation: Formation of a chelate; therapeutic administration of a chelating agent. Clinical Decision Guide (CDG): A new operational quantity introduced to guide physicians in considering the need for medical treatment. 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; a 30 d RBE-weighted absorbed dose value of 0.25 Gy-Eq (25 rad-Eq) for consideration of deterministic effects to bone marrow;

242 / GLOSSARY and a 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 FDA recommendations that potassium iodide be administered to adults >40 y of age if the projected dose to thyroid is ≥5 Gy, to adults 18 to 40 y of age if the projected dose is ≥0.1 Gy, and to pregnant or lactating women or persons <18 y of age if the projected dose is ≥0.05 Gy. committed dose: Integral of an internal dose rate over a specified period of time following an intake of a radionuclide by ingestion, inhalation, or dermal absorption. The commitment period of time (time of integration) is 50 y of age for workers and adults and to 70 y of age for children and pregnant women. committed effective dose (person-Sv): Most frequently the product of the mean effective dose for a population and the number of persons in the population, but, more precisely, and preferably, the sum of all individual effective doses in the population of concern. committed equivalent dose: Integral of an equivalent dose rate to an organ or tissue over a specified period of time following an intake of a radionuclide by ingestion, inhalation, or dermal absorption. For workers and adults, the commitment period of time (time of integration) is 50 y and to age 70 for children and pregnant women. contamination (radioactive): Radioactive material that is present in any substance or area or on any surface where its presence is unwanted or unexpected. curie (Ci): The special unit used previously for activity equal to 3.70 × 1010 becquerels (or disintegrations per second) (see activity and becquerel). debridement: The process of removing foreign objects and contaminated and devitalized tissue from or adjacent to a wound or lesion until surrounding healthy tissue is exposed. decontamination: Action taken to remove radionuclides from clothing and the external surfaces of the body, from rooms, building surfaces, equipment, and 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. depleted uranium (DU): Uranium with an isotopic content of <0.7 % 235 U; typically DU contains ~0.2 % 235U. deterministic effects: Harmful tissue reactions that occur in all individuals who receive greater than a threshold dose of radiation; the severity of the effect varies with the dose. Examples are radiation-induced cataracts (lens of the eye), radiation-induced erythema (skin), radiation induced pneumonitis (lungs), hematopoietic failure (bone marrow), hypothyroidism (thyroid), and gastrointestinal failure (gastrointestinal tract). diethylenetriamine pentaacetic acid (DTPA): Chelating substance that binds metal ions. DTPA is rapidly excreted from the body by the kidneys. disaster: An incident that overwhelms normal operations. Critical functions and infrastructure are unable to respond much less return to

GLOSSARY

/ 243

normal on their own, thus require regional or federal assistance (Farmer, 2006). dose: 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). Depending upon the context in which it used, the generic term dose may also refer to equivalent dose, effective dose or other dose-related quantities. dose coefficient: (1) For ingestion or inhalation of radionuclides, committed equivalent, or effective dose per unit activity intake; or (2) for external exposure to radionuclides in the environment, equivalent or effective dose rate per unit concentration in an environmental medium. dose rate: Dose per unit time; often expressed as an average over some time period (e.g., a year). Can refer to any dose quantity (e.g., absorbed, equivalent or effective dose). effective dose (E): The sum over specified organs and tissues of the products of the equivalent dose in an organ or tissue (HT) and the tissue weighting factor for that organ or tissue (wT):

E =

∑ wT HT .

(G.1)

T

The tissue weighting factors have been developed from a reference population of equal numbers of both males and females and a wide range of ages (ICRP, 1991a; 2007a). Effective dose (E) applies only to stochastic effects. The unit is joule per kilogram (J kg –1) with the special name sievert (Sv). emergency: A sudden, urgent, usually unforeseen occurrence or occasion requiring immediate action. emergency management: The process to achieve a full state of readiness. There are four phases to every disaster: preparedness, response, mitigation and recovery. Each phase takes planning, time and resources from a wide array of partners to ensure a coordinated response. The primary mission of the emergency management is to ensure the entity, institution and staff are prepared to respond to foreseeable disasters (Farmer, 2006). emergency operations center (EOC): A central command and control facility responsible for strategic overview of emergency response. equivalent dose (HT): Mean absorbed dose in a tissue or organ (DT,R) weighted by the radiation weighting factor (wR ) for the type and energy of radiation incident on the body:

HΤ =

∑ wR DT,R .

(G.2)

R

The SI unit of equivalent dose is joule per kilogram (J kg –1) with the special name sievert (Sv). 1 Sv = 1 J kg –1. ethylenediaminetetraacetic acid (EDTA): A chelating substance similar to diethylenetriamine pentaacetic acid (DTPA) in that it binds metal ions. EDTA is rapidly excreted from the body by the kidneys.

244 / GLOSSARY fallout: The radioactive material falling from the atmosphere to Earth’s surface after a radiological or nuclear incident, such as a weapons test or accident. family member: Any person who provides support and comfort to a patient on a regular basis and is considered by the patient as a member of their “family” whether by birth or marriage or by virtue of a close, loving relationship. gamma camera: An imaging device that displays the distribution of activity within a source such as the body. It records the quantity and distribution of photons emitted by the radioactive material in the area of interest. The gamma camera is the principal imaging device used in nuclear medicine. gamma radiation: Electromagnetic radiation (photons) emitted in de-excitation of atomic nuclei, and frequently occurring in decay of radionuclides. Also called gamma ray and sometimes shortened to gamma (e.g., gamma-emitting radionuclide) (see photon and x ray). gray (Gy): The special name for the SI unit of the quantities absorbed dose and air kerma. 1 Gy = 1 J kg–1. half-life (radioactive): The time in which one-half of the atoms (on average) of a particular radioactive substance disintegrate into another nuclear form (also called physical or radiological half-life). health physicist: An individual qualified by training and experience to be professionally engaged in the practice of health physics. health physics: The profession devoted to the protection of humans and their environment from potential radiation hazards, to the identification of potential beneficial effects of radiation, and to the assistance in the development of beneficial effects of ionizing and nonionizing radiation. improvised nuclear device (IND): A device designed by terrorists to produce a nuclear detonation. This includes stolen and subsequently modified nuclear weapons but does not include stockpiled weapons in the custody of the military. 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. internal contamination: Radioactive contamination of organs or tissues of an organism due to intakes of radionuclides (e.g., by ingestion, inhalation, through wounds, or dermal absorption). intravenous (IV): Within a vein. mean absorbed dose (DT): The mean absorbed dose in an organ or tissue, obtained by averaging absorbed doses at points in the organ or tissue. mitigation: Elimination of the threat or vulnerability or at least lessening the consequences or severity of the disaster (Farmer, 2006). Monte-Carlo simulation: Computation of a probability distribution of an output of a model on the basis of repeated calculations using random sampling of input variables from specified probability distributions. particle: A small volume of solid material, such as dust. preparedness: The process of anticipating potential vulnerabilities and developing comprehensive plans with all involved agencies, educating personnel on the plans, and practicing the plans (Farmer, 2006).

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radiation dose (or dose): A general term used when the context is not specific to a particular radiation dose quantity. When the context is specific, the name for the quantity is used (e.g., absorbed dose, equivalent dose, effective dose). radiation dose rate (or dose rate): The radiation dose delivered per unit time. radiation weighting factor: A factor used in radiation protection to place on a common scale, the biological effectiveness of different radiations when calculating equivalent dose (HT) (see equivalent dose). These factors are independent of the tissue or organ irradiated. radioactive contamination: Unintended and undesirable sources of radioactive materials deposited in the environment, research laboratories, hospitals, or other facilities and equipment and radioactive materials deposited on or in persons. radioactive decay: The spontaneous transformation of one nuclide into a different nuclide or into a different energy state of the same nuclide. The process results in a decrease, with time, of the number of the radioactive atoms in a sample. Decay generally involves the emission of alpha or beta particles from the nucleus of the decaying nuclide and possibly gamma rays from the newly formed nucleus. radioactivity: Property or characteristic of a unstable atomic nucleus to spontaneously transform with emission of energy in the form of ionizing radiation. radiological dispersal device (RDD): A device designed to spread radioactive material through a detonation of conventional explosive or other (non-nuclear) means. radiological triage: The process of sorting people involved in a radiological incident based on their risk of having significant radionuclide contamination or radiation dose. radionuclide: An unstable atom that transforms to a stable or unstable atom and in the process releases radiations. recovery: Returning to normal, or redefining normal, which is typically the most difficult and of longest duration (Farmer, 2006). relative biological effectiveness (RBE): Factor used to compare the biological effectiveness of absorbed doses from different types of ionizing radiation, determined experimentally. RBE is the ratio of the absorbed dose of a reference radiation to the absorbed dose of the radiation in question required to produce an identical biological effect in a particular experimental organism or tissue. response: How entities, institutions and people react to the threat (Farmer, 2006). roentgen (R): The special name for the unit of exposure. Exposure is a specific quantity of ionization (charge) produced by the absorption of gamma- or x-radiation energy in a specified mass of air under standard conditions. 1 R = 2.58 × 10–4 coulomb per kilogram (C kg–1). sealed source: Radioactive material encased in a capsule designed to prevent leakage or escape of the material. sievert (Sv): Special name for the SI unit of dose equivalent, equivalent dose, and effective dose. 1 Sv = 1 J kg–1. SI units: The International System of Units as defined by the General Conference of Weights and Measures in 1960. These units are generally

246 / GLOSSARY based on the meter/kilogram/second units, with special quantities for radiation including the becquerel, gray and sievert. specific activity: Activity of a radionuclide per unit mass of the radionuclide; also may refer to activity of a radionuclide per unit mass of material in which the radionuclide is dispersed. standard precautions: Previously known as Universal Precautions; an approach to infection control in which human blood and most human body fluids are treated as if infectious for human immunodeficiency virus, hepatitis B virus, or other blood-borne pathogens. stochastic: Of, pertaining to, or arising from chance; involving probability; random. stochastic risk: Probability for adverse effects in biological organisms for which the probability, but not the severity, is assumed to be a function of dose of ionizing radiation (or other contaminant) without threshold. survey: An evaluation of the presence of radiation or radioactive contamination under a specific set of conditions to determine actual or potential radiation hazards. survey meter: An instrument or device, usually portable, for monitoring the level of radiation or of radioactive contamination in an area or location. total effective dose: Integral of an effective dose rate over a specified finite period of time following an external radiation dose or an intake of a radionuclide by ingestion, inhalation, or dermal absorption or any combination of the three. x rays: Penetrating electromagnetic radiation having a range of wavelengths (energies) that are similar to those of gamma photons. X rays are usually produced by interaction of the electron field around certain nuclei or by the slowing down of energetic electrons. Once formed, there is no physical difference between gamma- and x-ray photons; however, there is a difference in their origin (see also bremsstrahlung).

Abbreviations and Acronyms AMAD AP ARS Ca-DTPA CDG CFR cpm dpm DU DTPA EDTA EOC FRMAC GI GM Gy-Eq HICS IND ICS IV KI LD50

PA RBE RDD REAC/TS SNS Zn-DTPA

activity median aerodynamic diameter antero-posterior acute radiation syndrome calcium diethylenetriamine pentaacetic acid Clinical Decision Guide Code of Federal Regulations counts per minute disintegrations per minute depleted uranium diethylenetriamine pentaacetic acid ethylenediaminetetraacetic acid (Edetate calcium disodium, calcium disodium Versonate®) emergency operations center Federal Radiological Monitoring and Assessment Center gastrointestinal Geiger-Mueller gray equivalent Hospital Incident Command System improvised nuclear device Incident Command System intravenous, intravenously potassium iodide lethal dose for causing death in 50 % of exposed persons by a particular time post-exposure (can also be defined for any other percentage of the population) postero-anterior relative biological effectiveness radiological dispersal device Radiation Emergency Assistance Center/Training Site Strategic National Stockpile zinc diethylenetriamine pentaacetic acid

247

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The NCRP The National Council on Radiation Protection and Measurements is a nonprofit corporation chartered by Congress in 1964 to: 1. Collect, analyze, develop and disseminate in the public interest information and recommendations about (a) protection against radiation and (b) radiation measurements, quantities and units, particularly those concerned with radiation protection. 2. Provide a means by which organizations concerned with the scientific and related aspects of radiation protection and of radiation quantities, units and measurements may cooperate for effective utilization of their combined resources, and to stimulate the work of such organizations. 3. Develop basic concepts about radiation quantities, units and measurements, about the application of these concepts, and about radiation protection. 4. Cooperate with the International Commission on Radiological Protection, the International Commission on Radiation Units and Measurements, and other national and international organizations, governmental and private, concerned with radiation quantities, units and measurements and with radiation protection. The Council is the successor to the unincorporated association of scientists known as the National Committee on Radiation Protection and Measurements and was formed to carry on the work begun by the Committee in 1929. The participants in the Council’s work are the Council members and members of scientific and administrative committees. Council members are selected solely on the basis of their scientific expertise and serve as individuals, not as representatives of any particular organization. The scientific committees, composed of experts having detailed knowledge and competence in the particular area of the committee's interest, draft proposed recommendations. These are then submitted to the full membership of the Council for careful review and approval before being published. The following comprise the current officers and membership of the Council:

Officers President Senior Vice President Secretary and Treasurer

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Thomas S. Tenforde Kenneth R. Kase David A. Schauer

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Members John F. Ahearne E. Stephen Amis, Jr. Sally A. Amundson Kimberly E. Applegate Benjamin R. Archer Stephen Balter Steven M. Becker Joel S. Bedford Mythreyi Bhargavan Eleanor A. Blakely William F. Blakely Wesley E. Bolch Thomas B. Borak Andre Bouville Leslie A. Braby James A. Brink Brooke R. Buddemeier Jerrold T. Bushberg John F. Cardella Charles E. Chambers Polly Y. Chang S.Y. Chen Lawrence L. Chi Mary E. Clark Michael L. Corradini Allen G. Croff Paul M. DeLuca Christine A. Donahue Stephen A. Feig Alan J. Fischman Patricia A. Fleming John R. Frazier Donald P. Frush

Ronald E. Goans Robert L. Goldberg Milton J. Guiberteau Raymond A. Guilmette Roger W. Harms Martin Hauer-Jensen Kathryn D. Held Roger W. Howell Hank C. Jenkins-Smith Timothy J. Jorgensen Kenneth R. Kase Ann R. Kennedy William E. Kennedy, Jr. David C. Kocher Ritsuko Komaki Amy Kronenberg Susan M. Langhorst John J. Lanza Edwin M. Leidholdt, Jr. James C. Lin Martha S. Linet Jill A. Lipoti Paul A. Locke Jay H. Lubin C. Douglas Maynard Debra McBaugh Ruth E. McBurney Charles W. Miller Donald L. Miller William H. Miller William F. Morgan Stephen V. Musolino David S. Myers Bruce A. Napier

Gregory A. Nelson Andrea K. Ng Carl J. Paperiello Terry C. Pellmar R. Julian Preston Kathryn H. Pryor Jerome C. Puskin Abram Recht Adela Salame-Alfie Beth A. Schueler J. Anthony Seibert Stephen M. Seltzer Edward A. Sickles Steven L. Simon Christopher G. Soares Michael G. Stabin Daniel J. Strom Tammy P. Taylor Thomas S. Tenforde Julie K. Timins Richard E. Toohey Elizabeth L. Travis Fong Y. Tsai Louis K. Wagner Chris G. Whipple Robert C. Whitcomb, Jr. Stuart C. White Gayle E. Woloschak Shiao Y. Woo Andrew J. Wyrobek X. George Xu R. Craig Yoder Marco A. Zaider

Distinguished Emeritus Members Warren K. Sinclair, President Emeritus; Charles B. Meinhold, President Emeritus S. James Adelstein, Honorary Vice President W. Roger Ney, Executive Director Emeritus William M. Beckner, Executive Director Emeritus Seymour Abrahamson Lynn R. Anspaugh John A. Auxier William J. Bair Harold L. Beck Bruce B. Boecker John D. Boice, Jr. Robert L. Brent Antone L. Brooks Randall S. Caswell J. Donald Cossairt James F. Crow Gerald D. Dodd Sarah S. Donaldson William P. Dornsife Keith F. Eckerman Thomas S. Ely R.J. Michael Fry Thomas F. Gesell

Ethel S. Gilbert Joel E. Gray Robert O. Gorson Arthur W. Guy Eric J. Hall Naomi H. Harley William R. Hendee F. Owen Hoffman Donald G. Jacobs Bernd Kahn Charles E. Land John B. Little Roger O. McClellan Barbara J. McNeil Fred A. Mettler, Jr. Kenneth L. Miller Dade W. Moeller A. Alan Moghissi

Wesley L. Nyborg John W. Poston, Sr. Andrew K. Poznanski Genevieve S. Roessler Marvin Rosenstein Lawrence N. Rothenberg Henry D. Royal Michael T. Ryan William J. Schull Roy E. Shore Paul Slovic John E. Till Lawrence W. Townsend Robert L. Ullrich Arthur C. Upton Richard J. Vetter F. Ward Whicker Susan D. Wiltshire Marvin C. Ziskin

266 / THE NCRP Lauriston S. Taylor Lecturers Charles E. Land (2010) Radiation Protection and Public Policy in an Uncertain World John D. Boice, Jr. (2009) Radiation Epidemiology: The Golden Age and Remaining Challenges Dade W. Moeller (2008) Radiation Standards, Dose/Risk Assessments, Public Interactions, and Yucca Mountain: Thinking Outside the Box Patricia W. Durbin (2007) The Quest for Therapeutic Actinide Chelators Robert L. Brent (2006) Fifty Years of Scientific Research: The Importance of Scholarship and the Influence of Politics and Controversy John B. Little (2005) Nontargeted Effects of Radiation: Implications for Low-Dose Exposures Abel J. Gonzalez (2004) Radiation Protection in the Aftermath of a Terrorist Attack Involving Exposure to Ionizing Radiation Charles B. Meinhold (2003) The Evolution of Radiation Protection: From Erythema to Genetic Risks to Risks of Cancer to ? R. Julian Preston (2002) Developing Mechanistic Data for Incorporation into Cancer Risk Assessment: Old Problems and New Approaches Wesley L. Nyborg (2001) Assuring the Safety of Medical Diagnostic Ultrasound S. James Adelstein (2000) Administered Radioactivity: Unde Venimus Quoque Imus Naomi H. Harley (1999) Back to Background Eric J. Hall (1998) From Chimney Sweeps to Astronauts: Cancer Risks in the Workplace William J. Bair (1997) Radionuclides in the Body: Meeting the Challenge! Seymour Abrahamson (1996) 70 Years of Radiation Genetics: Fruit Flies, Mice and Humans Albrecht Kellerer (1995) Certainty and Uncertainty in Radiation Protection R.J. Michael Fry (1994) Mice, Myths and Men Warren K. Sinclair (1993) Science, Radiation Protection and the NCRP Edward W. Webster (1992) Dose and Risk in Diagnostic Radiology: How Big? How Little? Victor P. Bond (1991) When is a Dose Not a Dose? J. Newell Stannard (1990) Radiation Protection and the Internal Emitter Saga Arthur C. Upton (1989) Radiobiology and Radiation Protection: The Past Century and Prospects for the Future Bo Lindell (1988) How Safe is Safe Enough? Seymour Jablon (1987) How to be Quantitative about Radiation Risk Estimates Herman P. Schwan (1986) Biological Effects of Non-ionizing Radiations: Cellular Properties and Interactions John H. Harley (1985) Truth (and Beauty) in Radiation Measurement Harald H. Rossi (1984) Limitation and Assessment in Radiation Protection Merril Eisenbud (1983) The Human Environment—Past, Present and Future Eugene L. Saenger (1982) Ethics, Trade-Offs and Medical Radiation James F. Crow (1981) How Well Can We Assess Genetic Risk? Not Very Harold O. Wyckoff (1980) From “Quantity of Radiation” and “Dose” to “Exposure” and “Absorbed Dose”—An Historical Review Hymer L. Friedell (1979) Radiation Protection—Concepts and Trade Offs Sir Edward Pochin (1978) Why be Quantitative about Radiation Risk Estimates?

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Herbert M. Parker (1977) The Squares of the Natural Numbers in Radiation Protection Currently, the following committees are actively engaged in formulating recommendations:

Program Area Committee 1: Basic Criteria, Epidemiology, Radiobiology, and Risk SC 1-15 Radiation Safety in NASA Lunar Missions’ SC 1-16 Uncertainties in the Estimation of Radiation Risks and Probability of Disease Causation SC 1-17 Second Cancers and Cardiopulmonary Effects After Radiotherapy SC 1-18 Use of Ionizing Radiation Screen Systems for Detection of Radioactive Materials that Could Represent a Threat to Homeland Security SC 1-19 Health Protection Issues Associated with Use of Active Detection Technology Security Systems for Detection of Radioactive Threat Materials SC 1-20 Biological Effectiveness of Photons as a Function of Energy

Program Area Committee 2: Operational Radiation Safety SC 2-5 Investigation of Radiological Incidents

Program Area Committee 3: Nuclear and Radiological Security and Safety Program Area Committee 4: Radiation Protection in Medicine SC 4-3 Diagnostic Reference Levels in Medical Imaging: Recommendations for Application in the United States SC 4-4 Risks of Ionizing Radiation to the Developing Embryo, Fetus and Nursing Infant

Program Area Committee 5: Environmental Radiation and Radioactive Waste Issues SC 5-1 Approach to Optimizing Decision Making for Late-Phase Recovery from Nuclear or Radiological Terrorism Incidents SC 64-22 Design of Effective Effluent and Environmental Monitoring Programs

Program Area Committee 6: Radiation Measurements and Dosimetry In recognition of its responsibility to facilitate and stimulate cooperation among organizations concerned with the scientific and related aspects of radiation protection and measurement, the Council has created a category of NCRP Collaborating Organizations. Organizations or groups of organizations that are national or international in scope and are concerned with scientific problems involving radiation quantities, units, measurements and effects, or radiation protection may be admitted to collaborating status by the Council. Collaborating Organizations provide a means by which NCRP can gain input into its activities from a wider segment of society. At the same time, the relationships with the Collaborating Organizations facilitate wider dissemination of information about the Council's activities, interests and concerns. Collaborating Organizations have the opportunity to comment on draft reports (at the time that these are submitted to the members of the Council). This is intended to capitalize on the fact that Collaborating Organizations are in an excellent position

268 / THE NCRP to both contribute to the identification of what needs to be treated in NCRP reports and to identify problems that might result from proposed recommendations. The present Collaborating Organizations with which NCRP maintains liaison are as follows: American Academy of Dermatology American Academy of Environmental Engineers American Academy of Health Physics American Academy of Orthopaedic Surgeons American Association of Physicists in Medicine American Bracytherapy Society American College of Cardiology American College of Medical Physics American College of Nuclear Physicians American College of Occupational and Environmental Medicine American College of Radiology American Conference of Governmental Industrial Hygienists American Dental Association American Industrial Hygiene Association American Institute of Ultrasound in Medicine American Medical Association American Nuclear Society American Pharmaceutical Association American Podiatric Medical Association American Public Health Association American Radium Society American Roentgen Ray Society American Society for Radiation Oncology American Society of Emergency Radiology American Society of Health-System Pharmacists American Society of Nuclear Cardiology American Society of Radiologic Technologists Association of Educators in Imaging and Radiological Sciences Association of University Radiologists Bioelectromagnetics Society Campus Radiation Safety Officers College of American Pathologists Conference of Radiation Control Program Directors, Inc. Council on Radionuclides and Radiopharmaceuticals Defense Threat Reduction Agency Electric Power Research Institute Federal Aviation Administration Federal Communications Commission Federal Emergency Management Agency Genetics Society of America Health Physics Society Institute of Electrical and Electronics Engineers, Inc. Institute of Nuclear Power Operations International Brotherhood of Electrical Workers International Society of Exposure Science National Aeronautics and Space Administration

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National Association of Environmental Professionals National Center for Environmental Health/Agency for Toxic Substances National Electrical Manufacturers Association National Institute for Occupational Safety and Health National Institute of Standards and Technology Nuclear Energy Institute Office of Science and Technology Policy Paper, Allied-Industrial, Chemical and Energy Workers International Union Product Stewardship Institute Radiation Research Society Radiological Society of North America Society for Cardiovascular Angiography and Interventions Society for Pediatric Radiology Society for Risk Analysis Society of Cardiovascular Computed Tomography Society of Chairmen of Academic Radiology Departments Society of Interventional Radiology Society of Nuclear Medicine Society of Radiologists in Ultrasound Society of Skeletal Radiology U.S. Air Force U.S. Army U.S. Coast Guard U.S. Department of Energy U.S. Department of Housing and Urban Development U.S. Department of Labor U.S. Department of Transportation U.S. Environmental Protection Agency U.S. Navy U.S. Nuclear Regulatory Commission U.S. Public Health Service Utility Workers Union of America NCRP has found its relationships with these organizations to be extremely valuable to continued progress in its program. Another aspect of the cooperative efforts of NCRP relates to the Special Liaison relationships established with various governmental organizations that have an interest in radiation protection and measurements. This liaison relationship provides: (1) an opportunity for participating organizations to designate an individual to provide liaison between the organization and NCRP; (2) that the individual designated will receive copies of draft NCRP reports (at the time that these are submitted to the members of the Council) with an invitation to comment, but not vote; and (3) that new NCRP efforts might be discussed with liaison individuals as appropriate, so that they might have an opportunity to make suggestions on new studies and related matters. The following organizations participate in the Special Liaison Program: Australian Radiation Laboratory Bundesamt fur Strahlenschutz (Germany) Canadian Association of Medical Radiation Technologists

270 / THE NCRP Canadian Nuclear Safety Commission Central Laboratory for Radiological Protection (Poland) China Institute for Radiation Protection Commissariat a l’Energie Atomique (France) Commonwealth Scientific Instrumentation Research Organization (Australia) European Commission Heads of the European Radiological Protection Competent Authorities Health Council of the Netherlands Health Protection Agency International Commission on Non-ionizing Radiation Protection International Commission on Radiation Units and Measurements International Commission on Radiological Protection International Radiation Protection Association Japanese Nuclear Safety Commission Japan Radiation Council Korea Institute of Nuclear Safety Russian Scientific Commission on Radiation Protection South African Forum for Radiation Protection World Association of Nuclear Operators World Health Organization, Radiation and Environmental Health NCRP values highly the participation of these organizations in the Special Liaison Program. The Council also benefits significantly from the relationships established pursuant to the Corporate Sponsor's Program. The program facilitates the interchange of information and ideas and corporate sponsors provide valuable fiscal support for the Council's program. This developing program currently includes the following Corporate Sponsors: 3M GE Healthcare Global Dosimetry Solutions, Inc. Landauer, Inc. Nuclear Energy Institute The Council's activities have been made possible by the voluntary contribution of time and effort by its members and participants and the generous support of the following organizations: 3M Health Physics Services Agfa Corporation Alfred P. Sloan Foundation Alliance of American Insurers American Academy of Dermatology American Academy of Health Physics American Academy of Oral and Maxillofacial Radiology American Association of Physicists in Medicine American Cancer Society American College of Medical Physics American College of Nuclear Physicians

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American College of Occupational and Environmental Medicine American College of Radiology American College of Radiology Foundation American Dental Association American Healthcare Radiology Administrators American Industrial Hygiene Association American Insurance Services Group American Medical Association American Nuclear Society American Osteopathic College of Radiology American Podiatric Medical Association American Public Health Association American Radium Society American Roentgen Ray Society American Society for Radiation Oncology American Society for Therapeutic Radiology and Oncology American Society of Radiologic Technologists American Veterinary Medical Association American Veterinary Radiology Society Association of Educators in Radiological Sciences, Inc. Association of University Radiologists Battelle Memorial Institute Canberra Industries, Inc. Chem Nuclear Systems Center for Devices and Radiological Health College of American Pathologists Committee on Interagency Radiation Research and Policy Coordination Commonwealth Edison Commonwealth of Pennsylvania Consolidated Edison Consumers Power Company Council on Radionuclides and Radiopharmaceuticals Defense Nuclear Agency Defense Threat Reduction Agency Duke Energy Corporation Eastman Kodak Company Edison Electric Institute Edward Mallinckrodt, Jr. Foundation EG&G Idaho, Inc. Electric Power Research Institute Electromagnetic Energy Association Federal Emergency Management Agency Florida Institute of Phosphate Research Florida Power Corporation Fuji Medical Systems, U.S.A., Inc. Genetics Society of America Global Dosimetry Solutions Health Effects Research Foundation (Japan) Health Physics Society ICN Biomedicals, Inc.

272 / THE NCRP Institute of Nuclear Power Operations James Picker Foundation Martin Marietta Corporation Motorola Foundation National Aeronautics and Space Administration National Association of Photographic Manufacturers National Cancer Institute National Electrical Manufacturers Association National Institute of Standards and Technology New York Power Authority Philips Medical Systems Picker International Public Service Electric and Gas Company Radiation Research Society Radiological Society of North America Richard Lounsbery Foundation Sandia National Laboratory Siemens Medical Systems, Inc. Society of Nuclear Medicine Society of Pediatric Radiology Southern California Edison Company U.S. Department of Energy U.S. Department of Labor U.S. Environmental Protection Agency U.S. Navy U.S. Nuclear Regulatory Commission Victoreen, Inc. Westinghouse Electric Corporation Initial funds for publication of NCRP reports were provided by a grant from the James Picker Foundation. NCRP seeks to promulgate information and recommendations based on leading scientific judgment on matters of radiation protection and measurement and to foster cooperation among organizations concerned with these matters. These efforts are intended to serve the public interest and the Council welcomes comments and suggestions on its reports or activities.

NCRP Publications NCRP publications can be obtained online in both soft- and hardcopy (downloadable PDF) formats at http://NCRPpublications.org. Professional societies can arrange for discounts for their members by contacting NCRP. Additional information on NCRP publications may be obtained from the NCRP website (http://NCRPonline.org) or by telephone (800-229-2652, ext. 25) and fax (301-907-8768). The mailing address is: NCRP Publications 7910 Woodmont Avenue Suite 400 Bethesda, MD 20814-3095 Abstracts of NCRP reports published since 1980, abstracts of all NCRP commentaries, and the text of all NCRP statements are available at the NCRP website. Currently available publications are listed below.

NCRP Reports No.

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25 27 30 32 35 36 37 38 40 41 42 44 46

Control and Removal of Radioactive Contamination in Laboratories (1951) Maximum Permissible Body Burdens and Maximum Permissible Concentrations of Radionuclides in Air and in Water for Occupational Exposure (1959) [includes Addendum 1 issued in August 1963] Measurement of Absorbed Dose of Neutrons, and of Mixtures of Neutrons and Gamma Rays (1961) Stopping Powers for Use with Cavity Chambers (1961) Safe Handling of Radioactive Materials (1964) Radiation Protection in Educational Institutions (1966) Dental X-Ray Protection (1970) Radiation Protection in Veterinary Medicine (1970) Precautions in the Management of Patients Who Have Received Therapeutic Amounts of Radionuclides (1970) Protection Against Neutron Radiation (1971) Protection Against Radiation from Brachytherapy Sources (1972) Specification of Gamma-Ray Brachytherapy Sources (1974) Radiological Factors Affecting Decision-Making in a Nuclear Attack (1974) Krypton-85 in the Atmosphere—Accumulation, Biological Significance, and Control Technology (1975) Alpha-Emitting Particles in Lungs (1975)

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274 / NCRP PUBLICATIONS 47 49 50 52 54 55 57 58 60 61 62 63 64 65 67 68 69 70 72 73 74 75 76

77 78 79 80 81 82

Tritium Measurement Techniques (1976) Structural Shielding Design and Evaluation for Medical Use of X Rays and Gamma Rays of Energies Up to 10 MeV (1976) Environmental Radiation Measurements (1976) Cesium-137 from the Environment to Man: Metabolism and Dose (1977) Medical Radiation Exposure of Pregnant and Potentially Pregnant Women (1977) Protection of the Thyroid Gland in the Event of Releases of Radioiodine (1977) Instrumentation and Monitoring Methods for Radiation Protection (1978) A Handbook of Radioactivity Measurements Procedures, 2nd ed. (1985) Physical, Chemical, and Biological Properties of Radiocerium Relevant to Radiation Protection Guidelines (1978) Radiation Safety Training Criteria for Industrial Radiography (1978) Tritium in the Environment (1979) Tritium and Other Radionuclide Labeled Organic Compounds Incorporated in Genetic Material (1979) Influence of Dose and Its Distribution in Time on Dose-Response Relationships for Low-LET Radiations (1980) Management of Persons Accidentally Contaminated with Radionuclides (1980) Radiofrequency Electromagnetic Fields—Properties, Quantities and Units, Biophysical Interaction, and Measurements (1981) Radiation Protection in Pediatric Radiology (1981) Dosimetry of X-Ray and Gamma-Ray Beams for Radiation Therapy in the Energy Range 10 keV to 50 MeV (1981) Nuclear Medicine—Factors Influencing the Choice and Use of Radionuclides in Diagnosis and Therapy (1982) Radiation Protection and Measurement for Low-Voltage Neutron Generators (1983) Protection in Nuclear Medicine and Ultrasound Diagnostic Procedures in Children (1983) Biological Effects of Ultrasound: Mechanisms and Clinical Implications (1983) Iodine-129: Evaluation of Releases from Nuclear Power Generation (1983) Radiological Assessment: Predicting the Transport, Bioaccumulation, and Uptake by Man of Radionuclides Released to the Environment (1984) Exposures from the Uranium Series with Emphasis on Radon and Its Daughters (1984) Evaluation of Occupational and Environmental Exposures to Radon and Radon Daughters in the United States (1984) Neutron Contamination from Medical Electron Accelerators (1984) Induction of Thyroid Cancer by Ionizing Radiation (1985) Carbon-14 in the Environment (1985) SI Units in Radiation Protection and Measurements (1985)

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83 The Experimental Basis for Absorbed-Dose Calculations in Medical Uses of Radionuclides (1985) 84 General Concepts for the Dosimetry of Internally Deposited Radionuclides (1985) 86 Biological Effects and Exposure Criteria for Radiofrequency Electromagnetic Fields (1986) 87 Use of Bioassay Procedures for Assessment of Internal Radionuclide Deposition (1987) 88 Radiation Alarms and Access Control Systems (1986) 89 Genetic Effects from Internally Deposited Radionuclides (1987) 90 Neptunium: Radiation Protection Guidelines (1988) 92 Public Radiation Exposure from Nuclear Power Generation in the United States (1987) 93 Ionizing Radiation Exposure of the Population of the United States (1987) 94 Exposure of the Population in the United States and Canada from Natural Background Radiation (1987) 95 Radiation Exposure of the U.S. Population from Consumer Products and Miscellaneous Sources (1987) 96 Comparative Carcinogenicity of Ionizing Radiation and Chemicals (1989) 97 Measurement of Radon and Radon Daughters in Air (1988) 99 Quality Assurance for Diagnostic Imaging (1988) 100 Exposure of the U.S. Population from Diagnostic Medical Radiation (1989) 101 Exposure of the U.S. Population from Occupational Radiation (1989) 102 Medical X-Ray, Electron Beam and Gamma-Ray Protection for Energies Up to 50 MeV (Equipment Design, Performance and Use) (1989) 103 Control of Radon in Houses (1989) 104 The Relative Biological Effectiveness of Radiations of Different Quality (1990) 105 Radiation Protection for Medical and Allied Health Personnel (1989) 106 Limit for Exposure to “Hot Particles” on the Skin (1989) 107 Implementation of the Principle of As Low As Reasonably Achievable (ALARA) for Medical and Dental Personnel (1990) 108 Conceptual Basis for Calculations of Absorbed-Dose Distributions (1991) 109 Effects of Ionizing Radiation on Aquatic Organisms (1991) 110 Some Aspects of Strontium Radiobiology (1991) 111 Developing Radiation Emergency Plans for Academic, Medical or Industrial Facilities (1991) 112 Calibration of Survey Instruments Used in Radiation Protection for the Assessment of Ionizing Radiation Fields and Radioactive Surface Contamination (1991) 113 Exposure Criteria for Medical Diagnostic Ultrasound: I. Criteria Based on Thermal Mechanisms (1992) 114 Maintaining Radiation Protection Records (1992) 115 Risk Estimates for Radiation Protection (1993) 116 Limitation of Exposure to Ionizing Radiation (1993)

276 / NCRP PUBLICATIONS 117 Research Needs for Radiation Protection (1993) 118 Radiation Protection in the Mineral Extraction Industry (1993) 119 A Practical Guide to the Determination of Human Exposure to Radiofrequency Fields (1993) 120 Dose Control at Nuclear Power Plants (1994) 121 Principles and Application of Collective Dose in Radiation Protection (1995) 122 Use of Personal Monitors to Estimate Effective Dose Equivalent and Effective Dose to Workers for External Exposure to Low-LET Radiation (1995) 123 Screening Models for Releases of Radionuclides to Atmosphere, Surface Water, and Ground (1996) 124 Sources and Magnitude of Occupational and Public Exposures from Nuclear Medicine Procedures (1996) 125 Deposition, Retention and Dosimetry of Inhaled Radioactive Substances (1997) 126 Uncertainties in Fatal Cancer Risk Estimates Used in Radiation Protection (1997) 127 Operational Radiation Safety Program (1998) 128 Radionuclide Exposure of the Embryo/Fetus (1998) 129 Recommended Screening Limits for Contaminated Surface Soil and Review of Factors Relevant to Site-Specific Studies (1999) 130 Biological Effects and Exposure Limits for “Hot Particles” (1999) 131 Scientific Basis for Evaluating the Risks to Populations from Space Applications of Plutonium (2001) 132 Radiation Protection Guidance for Activities in Low-Earth Orbit (2000) 133 Radiation Protection for Procedures Performed Outside the Radiology Department (2000) 134 Operational Radiation Safety Training (2000) 135 Liver Cancer Risk from Internally-Deposited Radionuclides (2001) 136 Evaluation of the Linear-Nonthreshold Dose-Response Model for Ionizing Radiation (2001) 137 Fluence-Based and Microdosimetric Event-Based Methods for Radiation Protection in Space (2001) 138 Management of Terrorist Events Involving Radioactive Material (2001) 139 Risk-Based Classification of Radioactive and Hazardous Chemical Wastes (2002) 140 Exposure Criteria for Medical Diagnostic Ultrasound: II. Criteria Based on all Known Mechanisms (2002) 141 Managing Potentially Radioactive Scrap Metal (2002) 142 Operational Radiation Safety Program for Astronauts in Low-Earth Orbit: A Basic Framework (2002) 143 Management Techniques for Laboratories and Other Small Institutional Generators to Minimize Off-Site Disposal of Low-Level Radioactive Waste (2003) 144 Radiation Protection for Particle Accelerator Facilities (2003) 145 Radiation Protection in Dentistry (2003)

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146 Approaches to Risk Management in Remediation of Radioactively Contaminated Sites (2004) 147 Structural Shielding Design for Medical X-Ray Imaging Facilities (2004) 148 Radiation Protection in Veterinary Medicine (2004) 149 A Guide to Mammography and Other Breast Imaging Procedures (2004) 150 Extrapolation of Radiation-Induced Cancer Risks from Nonhuman Experimental Systems to Humans (2005) 151 Structural Shielding Design and Evaluation for Megavoltage X- and Gamma-Ray Radiotherapy Facilities (2005) 152 Performance Assessment of Near-Surface Facilities for Disposal of Low-Level Radioactive Waste (2005) 153 Information Needed to Make Radiation Protection Recommendations for Space Missions Beyond Low-Earth Orbit (2006) 154 Cesium-137 in the Environment: Radioecology and Approaches to Assessment and Management (2006) 155 Management of Radionuclide Therapy Patients (2006) 156 Development of a Biokinetic Model for Radionuclide-Contaminated Wounds and Procedures for Their Assessment, Dosimetry and Treatment (2006) 157 Radiation Protection in Educational Institutions (2007) 158 Uncertainties in the Measurement and Dosimetry of External Radiation (2007) 159 Risk to the Thyroid from Ionizing Radiation (2008) 160 Ionizing Radiation Exposure of the Population of the United States (2009) 161 Management of Persons Contaminated with Radionuclides (2008) 162 Self Assessment of Radiation-Safety Programs (2009) 163 Radiation Dose Reconstruction: Principles and Practices (2009) 164 Uncertainties in Internal Radiation Dose Assessment (2009) 165 Responding to a Radiological or Nuclear Terrorism Incident: A Guide for Decision Makers (2010) 166 Population Monitoring and Radionuclide Decorporation Following a Radiological or Nuclear Incident (2010) 167 Potential Impact of Individual Genetic Susceptibility and Previous Radiation Exposure on Radiation Risk for Astronauts (2010) 168 Radiation Dose Management for Fluoroscopically-Guided Interventional Medical Procedures (2010) Binders for NCRP reports are available. Two sizes make it possible to collect into small binders the “old series” of reports (NCRP Reports Nos. 8–30) and into large binders the more recent publications (NCRP Reports Nos. 32–167, 168). Each binder will accommodate from five to seven reports. The binders carry the identification “NCRP Reports” and come with label holders which permit the user to attach labels showing the reports contained in each binder. The following bound sets of NCRP reports are also available: Volume I. NCRP Reports Nos. 8, 22 Volume II. NCRP Reports Nos. 23, 25, 27, 30

278 / NCRP PUBLICATIONS Volume III. NCRP Reports Nos. 32, 35, 36, 37 Volume IV. NCRP Reports Nos. 38, 40, 41 Volume V. NCRP Reports Nos. 42, 44, 46 Volume VI. NCRP Reports Nos. 47, 49, 50, 51 Volume VII. NCRP Reports Nos. 52, 53, 54, 55, 57 Volume VIII. NCRP Report No. 58 Volume IX. NCRP Reports Nos. 59, 60, 61, 62, 63 Volume X. NCRP Reports Nos. 64, 65, 66, 67 Volume XI. NCRP Reports Nos. 68, 69, 70, 71, 72 Volume XII. NCRP Reports Nos. 73, 74, 75, 76 Volume XIII. NCRP Reports Nos. 77, 78, 79, 80 Volume XIV. NCRP Reports Nos. 81, 82, 83, 84, 85 Volume XV. NCRP Reports Nos. 86, 87, 88, 89 Volume XVI. NCRP Reports Nos. 90, 91, 92, 93 Volume XVII. NCRP Reports Nos. 94, 95, 96, 97 Volume XVIII. NCRP Reports Nos. 98, 99, 100 Volume XIX. NCRP Reports Nos. 101, 102, 103, 104 Volume XX. NCRP Reports Nos. 105, 106, 107, 108 Volume XXI. NCRP Reports Nos. 109, 110, 111 Volume XXII. NCRP Reports Nos. 112, 113, 114 Volume XXIII. NCRP Reports Nos. 115, 116, 117, 118 Volume XXIV. NCRP Reports Nos. 119, 120, 121, 122 Volume XXV. NCRP Report No. 123I and 123II Volume XXVI. NCRP Reports Nos. 124, 125, 126, 127 Volume XXVII. NCRP Reports Nos. 128, 129, 130 Volume XXVIII. NCRP Reports Nos. 131, 132, 133 Volume XXIX. NCRP Reports Nos. 134, 135, 136, 137 Volume XXX. NCRP Reports Nos. 138, 139 Volume XXXI. NCRP Report No. 140 Volume XXXII. NCRP Reports Nos. 141, 142, 143 Volume XXXIII. NCRP Report No. 144 Volume XXXIV. NCRP Reports Nos. 145, 146, 147 Volume XXXV. NCRP Reports Nos. 148, 149 Volume XXXVI. NCRP Reports Nos. 150, 151, 152 Volume XXXVII, NCRP Reports Nos. 153, 154, 155 Volume XXXVIII, NCRP Reports Nos. 156, 157, 158 Volume XXXIX, NCRP Reports Nos. 159, 160 Volume XL. NCRP Report No. 161 (Vol I and II) Volume XLI. NCRP Reports Nos. 162, 163 (Titles of the individual reports contained in each volume are given previously.)

NCRP Commentaries No.

Title 1

Krypton-85 in the Atmosphere—With Specific Reference to the Public Health Significance of the Proposed Controlled Release at Three Mile Island (1980)

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Guidelines for the Release of Waste Water from Nuclear Facilities with Special Reference to the Public Health Significance of the Proposed Release of Treated Waste Waters at Three Mile Island (1987) Review of the Publication, Living Without Landfills (1989) Radon Exposure of the U.S. Population—Status of the Problem (1991) Misadministration of Radioactive Material in Medicine—Scientific Background (1991) Uncertainty in NCRP Screening Models Relating to Atmospheric Transport, Deposition and Uptake by Humans (1993) Considerations Regarding the Unintended Radiation Exposure of the Embryo, Fetus or Nursing Child (1994) Advising the Public about Radiation Emergencies: A Document for Public Comment (1994) Dose Limits for Individuals Who Receive Exposure from Radionuclide Therapy Patients (1995) Radiation Exposure and High-Altitude Flight (1995) An Introduction to Efficacy in Diagnostic Radiology and Nuclear Medicine (Justification of Medical Radiation Exposure) (1995) A Guide for Uncertainty Analysis in Dose and Risk Assessments Related to Environmental Contamination (1996) Evaluating the Reliability of Biokinetic and Dosimetric Models and Parameters Used to Assess Individual Doses for Risk Assessment Purposes (1998) Screening of Humans for Security Purposes Using Ionizing Radiation Scanning Systems (2003) Pulsed Fast Neutron Analysis System Used in Security Surveillance (2003) Biological Effects of Modulated Radiofrequency Fields (2003) Key Elements of Preparing Emergency Responders for Nuclear and Radiological Terrorism (2005) Radiation Protection and Measurement Issues Related to Cargo Scanning with Accelerator-Produced High-Energy X Rays (2007)

Proceedings of the Annual Meeting No.

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Perceptions of Risk, Proceedings of the Fifteenth Annual Meeting held on March 14-15, 1979 (including Taylor Lecture No. 3) (1980) Critical Issues in Setting Radiation Dose Limits, Proceedings of the Seventeenth Annual Meeting held on April 8-9, 1981 (including Taylor Lecture No. 5) (1982) Radiation Protection and New Medical Diagnostic Approaches, Proceedings of the Eighteenth Annual Meeting held on April 6-7, 1982 (including Taylor Lecture No. 6) (1983) Environmental Radioactivity, Proceedings of the Nineteenth Annual Meeting held on April 6-7, 1983 (including Taylor Lecture No. 7) (1983) Some Issues Important in Developing Basic Radiation Protection Recommendations, Proceedings of the Twentieth Annual Meeting held on April 4-5, 1984 (including Taylor Lecture No. 8) (1985)

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Radioactive Waste, Proceedings of the Twenty-First Annual Meeting held on April 3-4, 1985 (including Taylor Lecture No. 9)(1986) Nonionizing Electromagnetic Radiations and Ultrasound, Proceedings of the Twenty-Second Annual Meeting held on April 2-3, 1986 (including Taylor Lecture No. 10) (1988) New Dosimetry at Hiroshima and Nagasaki and Its Implications for Risk Estimates, Proceedings of the Twenty-Third Annual Meeting held on April 8-9, 1987 (including Taylor Lecture No. 11) (1988) Radon, Proceedings of the Twenty-Fourth Annual Meeting held on March 30-31, 1988 (including Taylor Lecture No. 12) (1989) Radiation Protection Today—The NCRP at Sixty Years, Proceedings of the Twenty-Fifth Annual Meeting held on April 5-6, 1989 (including Taylor Lecture No. 13) (1990) Health and Ecological Implications of Radioactively Contaminated Environments, Proceedings of the Twenty-Sixth Annual Meeting held on April 4-5, 1990 (including Taylor Lecture No. 14) (1991) Genes, Cancer and Radiation Protection, Proceedings of the Twenty-Seventh Annual Meeting held on April 3-4, 1991 (including Taylor Lecture No. 15) (1992) Radiation Protection in Medicine, Proceedings of the Twenty-Eighth Annual Meeting held on April 1-2, 1992 (including Taylor Lecture No. 16) (1993) Radiation Science and Societal Decision Making, Proceedings of the Twenty-Ninth Annual Meeting held on April 7-8, 1993 (including Taylor Lecture No. 17) (1994) Extremely-Low-Frequency Electromagnetic Fields: Issues in Biological Effects and Public Health, Proceedings of the Thirtieth Annual Meeting held on April 6-7, 1994 (not published). Environmental Dose Reconstruction and Risk Implications, Proceedings of the Thirty-First Annual Meeting held on April 12-13, 1995 (including Taylor Lecture No. 19) (1996) Implications of New Data on Radiation Cancer Risk, Proceedings of the Thirty-Second Annual Meeting held on April 3-4, 1996 (including Taylor Lecture No. 20) (1997) The Effects of Pre- and Postconception Exposure to Radiation, Proceedings of the Thirty-Third Annual Meeting held on April 2-3, 1997, Teratology 59, 181–317 (1999) Cosmic Radiation Exposure of Airline Crews, Passengers and Astronauts, Proceedings of the Thirty-Fourth Annual Meeting held on April 1-2, 1998, Health Phys. 79, 466–613 (2000) Radiation Protection in Medicine: Contemporary Issues, Proceedings of the Thirty-Fifth Annual Meeting held on April 7-8, 1999 (including Taylor Lecture No. 23) (1999) Ionizing Radiation Science and Protection in the 21st Century, Proceedings of the Thirty-Sixth Annual Meeting held on April 5-6, 2000, Health Phys. 80, 317–402 (2001) Fallout from Atmospheric Nuclear Tests—Impact on Science and Society, Proceedings of the Thirty-Seventh Annual Meeting held on April 4-5, 2001, Health Phys. 82, 573–748 (2002)

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Where the New Biology Meets Epidemiology: Impact on Radiation Risk Estimates, Proceedings of the Thirty-Eighth Annual Meeting held on April 10-11, 2002, Health Phys. 85, 1–108 (2003) Radiation Protection at the Beginning of the 21st Century–A Look Forward, Proceedings of the Thirty-Ninth Annual Meeting held on April 9–10, 2003, Health Phys. 87, 237–319 (2004) Advances in Consequence Management for Radiological Terrorism Events, Proceedings of the Fortieth Annual Meeting held on April 14–15, 2004, Health Phys. 89, 415–588 (2005) Managing the Disposition of Low-Activity Radioactive Materials, Proceedings of the Forty-First Annual Meeting held on March 30–31, 2005, Health Phys. 91, 413–536 (2006) Chernobyl at Twenty, Proceedings of the Forty-Second Annual Meeting held on April 3–4, 2006, Health Phys. 93, 345–595 (2007) Advances in Radiation Protection in Medicine, Proceedings of the Forty-Third Annual Meeting held on April 16-17, 2007, Health Phys. 95, 461–686 (2008) Low Dose and Low Dose-Rate Radiation Effects and Models, Proceedings of the Forty-Fourth Annual Meeting held on April 14–15, 2008, Health Phys. 97, 373–541 (2009)

Lauriston S. Taylor Lectures No.

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The Squares of the Natural Numbers in Radiation Protection by Herbert M. Parker (1977) Why be Quantitative about Radiation Risk Estimates? by Sir Edward Pochin (1978) Radiation Protection—Concepts and Trade Offs by Hymer L. Friedell (1979) [available also in Perceptions of Risk, see above] From “Quantity of Radiation” and “Dose” to “Exposure” and “Absorbed Dose”—An Historical Review by Harold O. Wyckoff (1980) How Well Can We Assess Genetic Risk? Not Very by James F. Crow (1981) [available also in Critical Issues in Setting Radiation Dose Limits, see above] Ethics, Trade-offs and Medical Radiation by Eugene L. Saenger (1982) [available also in Radiation Protection and New Medical Diagnostic Approaches, see above] The Human Environment—Past, Present and Future by Merril Eisenbud (1983) [available also in Environmental Radioactivity, see above] Limitation and Assessment in Radiation Protection by Harald H. Rossi (1984) [available also in Some Issues Important in Developing Basic Radiation Protection Recommendations, see above] Truth (and Beauty) in Radiation Measurement by John H. Harley (1985) [available also in Radioactive Waste, see above] Biological Effects of Non-ionizing Radiations: Cellular Properties and Interactions by Herman P. Schwan (1987) [available also in Nonionizing Electromagnetic Radiations and Ultrasound, see above]

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How to be Quantitative about Radiation Risk Estimates by Seymour Jablon (1988) [available also in New Dosimetry at Hiroshima and Nagasaki and its Implications for Risk Estimates, see above] How Safe is Safe Enough? by Bo Lindell (1988) [available also in Radon, see above] Radiobiology and Radiation Protection: The Past Century and Prospects for the Future by Arthur C. Upton (1989) [available also in Radiation Protection Today, see above] Radiation Protection and the Internal Emitter Saga by J. Newell Stannard (1990) [available also in Health and Ecological Implications of Radioactively Contaminated Environments, see above] When is a Dose Not a Dose? by Victor P. Bond (1992) [available also in Genes, Cancer and Radiation Protection, see above] Dose and Risk in Diagnostic Radiology: How Big? How Little? by Edward W. Webster (1992) [available also in Radiation Protection in Medicine, see above] Science, Radiation Protection and the NCRP by Warren K. Sinclair (1993) [available also in Radiation Science and Societal Decision Making, see above] Mice, Myths and Men by R.J. Michael Fry (1995) Certainty and Uncertainty in Radiation Research by Albrecht M. Kellerer. Health Phys. 69, 446–453 (1995) 70 Years of Radiation Genetics: Fruit Flies, Mice and Humans by Seymour Abrahamson. Health Phys. 71, 624–633 (1996) Radionuclides in the Body: Meeting the Challenge by William J. Bair. Health Phys. 73, 423–432 (1997) From Chimney Sweeps to Astronauts: Cancer Risks in the Work Place by Eric J. Hall. Health Phys. 75, 357–366 (1998) Back to Background: Natural Radiation and Radioactivity Exposed by Naomi H. Harley. Health Phys. 79, 121–128 (2000) Administered Radioactivity: Unde Venimus Quoque Imus by S. James Adelstein. Health Phys. 80, 317–324 (2001) Assuring the Safety of Medical Diagnostic Ultrasound by Wesley L. Nyborg. Health Phys. 82, 578–587 (2002) Developing Mechanistic Data for Incorporation into Cancer and Genetic Risk Assessments: Old Problems and New Approaches by R. Julian Preston. Health Phys. 85, 4–12 (2003) The Evolution of Radiation Protection–From Erythema to Genetic Risks to Risks of Cancer to ? by Charles B. Meinhold, Health Phys. 87, 240–248 (2004) Radiation Protection in the Aftermath of a Terrorist Attack Involving Exposure to Ionizing Radiation by Abel J. Gonzalez, Health Phys. 89, 418–446 (2005) Nontargeted Effects of Radiation: Implications for Low Dose Exposures by John B. Little, Health Phys. 91, 416–426 (2006) Fifty Years of Scientific Research: The Importance of Scholarship and the Influence of Politics and Controversy by Robert L. Brent, Health Phys. 93, 348–379 (2007) The Quest for Therapeutic Actinide Chelators by Patricia W. Durbin, Health Phys. 95, 465–492 (2008)

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Yucca Mountain Radiation Standards, Dose/Risk Assessments, Thinking Outside the Box, Evaluations, and Recommendations by Dade W. Moeller, Health Phys. 97, 376–391 (2009)

Symposium Proceedings No.

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The Control of Exposure of the Public to Ionizing Radiation in the Event of Accident or Attack, Proceedings of a Symposium held April 27-29, 1981 (1982) Radioactive and Mixed Waste—Risk as a Basis for Waste Classification, Proceedings of a Symposium held November 9, 1994 (1995) Acceptability of Risk from Radiation—Application to Human Space Flight, Proceedings of a Symposium held May 29, 1996 (1997) 21st Century Biodosimetry: Quantifying the Past and Predicting the Future, Proceedings of a Symposium held February 22, 2001, Radiat. Prot. Dosim. 97(1), (2001) National Conference on Dose Reduction in CT, with an Emphasis on Pediatric Patients, Summary of a Symposium held November 6-7, 2002, Am. J. Roentgenol. 181(2), 321–339 (2003)

NCRP Statements No.

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“Blood Counts, Statement of the National Committee on Radiation Protection,” Radiology 63, 428 (1954) “Statements on Maximum Permissible Dose from Television Receivers and Maximum Permissible Dose to the Skin of the Whole Body,” Am. J. Roentgenol., Radium Ther. and Nucl. Med. 84, 152 (1960) and Radiology 75, 122 (1960) X-Ray Protection Standards for Home Television Receivers, Interim Statement of the National Council on Radiation Protection and Measurements (1968) Specification of Units of Natural Uranium and Natural Thorium, Statement of the National Council on Radiation Protection and Measurements (1973) NCRP Statement on Dose Limit for Neutrons (1980) Control of Air Emissions of Radionuclides (1984) The Probability That a Particular Malignancy May Have Been Caused by a Specified Irradiation (1992) The Application of ALARA for Occupational Exposures (1999) Extension of the Skin Dose Limit for Hot Particles to Other External Sources of Skin Irradiation (2001) Recent Applications of the NCRP Public Dose Limit Recommendation for Ionizing Radiation (2004)

284 / NCRP PUBLICATIONS Other Documents The following documents were published outside of the NCRP report, commentary and statement series: Somatic Radiation Dose for the General Population, Report of the Ad Hoc Committee of the National Council on Radiation Protection and Measurements, 6 May 1959, Science 131 (3399), February 19, 482–486 (1960) Dose Effect Modifying Factors in Radiation Protection, Report of Subcommittee M-4 (Relative Biological Effectiveness) of the National Council on Radiation Protection and Measurements, Report BNL 50073 (T-471) (1967) Brookhaven National Laboratory (National Technical Information Service, Springfield, Virginia) Residential Radon Exposure and Lung Cancer Risk: Commentary on Cohen's County-Based Study, Health Phys. 87(6), 656–658 (2004)